ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 152

Document Title:

Site Profile for the West Valley 

Demonstration Project

Document Number: ORAUT-TKBS-0057

Revision: 00

Effective Date: 06/22/2007

Type of Document: TBD

Supersedes: None

Subject Expert(s): Robert E. Burns, Jr.

Site Expert(s): N/A

Approval: Signature on File Approval Date: 05/31/2007

Robert E. Burns, Jr., Document Owner

Concurrence: Signature on File Concurrence Date: 05/31/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 06/01/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 05/31/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 06/22/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 2 of 152

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

06/22/2007 00 New approved Site Profile for the West Valley Demonstration Project. 

Incorporates formal internal and NIOSH review comments. There is 

an increase in assigned dose and a PER is required. Training 

required: As determined by the Task Manager. Initiated by Robert 

E. Burns, Jr.
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 3 of 152

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations .............................................................................................................

PAGE

13

1.0 Introduction ............................................................................................................................17

1.1 Purpose......................................................................................................................18

1.2 Scope.........................................................................................................................18

2.0 Site Description......................................................................................................................18

2.1 Site History.................................................................................................................18

2.2 Reprocessing Plant Operations Era, 1965 to 1973 .....................................................20

2.2.1 Plant and Process Description ..................................................................... 20

2.2.2 Radiation Protection Program ...................................................................... 21

2.2.3 Special Considerations for the Operations Era............................................. 22

2.2.4 Chronic Sources of Personnel Exposures .................................................... 23

 2.2.4.1 Plant Ventilation Systems................................................................ 23

 2.2.4.2 Acid Recovery System.................................................................... 24

 2.2.4.3 Fuel Receiving and Storage Facility ................................................ 25

2.2.5 Miscellaneous Sources of Personnel Exposures.......................................... 25

2.2.6 Personnel Exposures Associated with Specific Tasks or 

Duties........................................................................................................... 26

 2.2.6.1 Maintenance Workers...................................................................... 26

 2.2.6.2 Operations....................................................................................... 30

 2.2.6.3 Ventilation System Filter Changeouts.............................................. 31

 2.2.6.4 Overall Workforce........................................................................... 35

2.3 Reprocessing Plant Residual Radiation Era, 1974 to February 25, 

1982 ...........................................................................................................................38

2.4 Reprocessing Plant DOE Era, February 26, 1982, to present..................................... 41

2.5 State-Licensed (Commercial) Waste Disposal Area, 1963 to 1983............................. 44

2.6 High-Level Waste Disposal Area, 1966 to present......................................................46

2.7 Isotopic Information for Plant Areas............................................................................47

2.7.1 Operations and Residual Radiation Eras, 1966 to February 

25, 1982....................................................................................................... 47

2.7.2 DOE Era, February 26, 1982, to present ...................................................... 49

3.0 Occupational Medical Dose.................................................................................................... 50

4.0 Occupational Environmental Dose ......................................................................................... 51

4.1 Inhalation and submersion Dose ................................................................................52

4.1.1 Stack Releases ............................................................................................ 52

 4.1.1.1 Particulates..................................................................................... 52

 4.1.1.2 Krypton-85...................................................................................... 54

 4.1.1.3 Iodine-131 ....................................................................................... 54

 4.1.1.4 Tritium............................................................................................. 55

4.1.2 Ground Level Releases................................................................................ 56

4.2 Ambient Dose.............................................................................................................57

4.3 Summary of Exposures from Environmental Sources.................................................58

5.0 Internal Dosimetry..................................................................................................................59
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5.1 Internal Dose Control Program ...................................................................................59

5.1.1 Operations and Residual Radiation Eras 1966 – February 

25, 1982....................................................................................................... 59

 5.1.1.1 Nasal Swabs ................................................................................... 60

 5.1.1.2 Chest Counter................................................................................. 60

5.1.2 DOE Era February 26, 1982 – present......................................................... 60

5.2 Bioassay program.......................................................................................................61

5.2.1 Operations and Residual Radiation Eras 1966 – February 

25, 1982....................................................................................................... 61

 5.2.1.1 Bioassay Records........................................................................... 63

5.2.2 DOE Era February 26, 1982 - present.......................................................... 63

 5.2.2.1 Routine Bioassay ............................................................................ 63

 5.2.2.2 Special Bioassay............................................................................. 64

5.2.3 Summary ..................................................................................................... 64

5.3 In Vitro Analyses.........................................................................................................65

5.3.1 Operations and Residual Radiation Eras 1966 – February 

25, 1982....................................................................................................... 65

5.3.2 DOE Era February 26, 1982 - present.......................................................... 68

5.4 In Vivo Analyses.........................................................................................................69

5.4.1 Operations and Residual Radiation Eras...................................................... 69

 5.4.1.1 Chest Counter ................................................................................ 69

 5.4.1.2 Commercial Mobile Whole-Body Counting ...................................... 70

 5.4.1.3 New York University Medical Center Phoswich 

 Counter ........................................................................................... 72

5.4.2 U.S. Department of Energy Era.................................................................... 72

5.5 Interpretation of Bioassay Measurements for West Valley Workers............................ 73

6.0 External Dosimetry................................................................................................................. 75

6.1 Basis of Comparison ..................................................................................................75

6.2 Site Historical Administrative Practices.......................................................................75

6.2.1 Site Access Requirements ........................................................................... 75

6.2.2 Personnel Exposure Control ........................................................................ 77

6.2.3 Evaluation and Recording of Personnel Exposures...................................... 78

 6.2.3.1 Dose Recording Practices ............................................................... 78

 6.2.3.2 Nuclear-Chicago Film Badge Exposure Reports............................. 80

 6.2.3.3 Landauer Radiation Dosimetry Reports........................................... 81

6.3 Site Dosimetry Technology.........................................................................................81

6.3.1 Whole-Body Monitoring................................................................................ 81

6.3.2 Extremity Monitoring .................................................................................... 81

6.4 Workplace Radiation Fields ........................................................................................82

6.4.1 Beta and Photon .......................................................................................... 82

6.4.2 Neutron........................................................................................................ 82

6.4.3 Summary ..................................................................................................... 82

6.5 Missed Dose...............................................................................................................83

6.6 Uncertainty .................................................................................................................84

7.0 Attributions and Annotations .................................................................................................. 84

References ........................................................................................................................................87
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 5 of 152

ATTACHMENT A, DESCRIPTIONS OF OPERATIONS AREAS AND SYSTEMS 

WITH EMPHASIS ON RADIOLOGICAL CONDITIONS..................................... 100

A1.0 Acid Recovery Cell............................................................................................................... 106

A1.1 Radiological Conditions ............................................................................................106

A1.1.1 External Exposure Data ............................................................................. 106

A2.0 Acid Recovery Pump Room ................................................................................................. 106

A2.1 Radiological Conditions ............................................................................................107

A2.1.1 External Exposure Data ............................................................................. 107

A2.1.2 Removable Contamination Data................................................................. 107

A3.0 Analytical Aisle, Laboratories, and North Operating Aisle..................................................... 107

A3.1 Radiological Conditions ............................................................................................108

A3.1.1 External Exposure Data ............................................................................. 108

A3.1.2 Removable Contamination Data................................................................. 109

A4.0 Cell Access Aisle ................................................................................................................. 109

A4.1 Radiological Conditions ............................................................................................109

A4.1.1 Removable Contamination Data................................................................. 109

A4.1.2 Airborne Contamination Data ..................................................................... 109

A5.0 Cell Crane Rooms................................................................................................................ 109

A5.1 Impact of the Thorex Campaign on Airborne Activity Levels in the Cell 

Crane Rooms ...........................................................................................................110

A5.2 General Purpose Cell Crane Room and Extension................................................... 110

A5.2.1 Radiological Conditions.............................................................................. 110

A5.2.2 External Exposure Data ............................................................................. 110

A5.2.3 Removable Contamination Data................................................................. 111

A5.2.4 Airborne Contamination Data ..................................................................... 111

A5.3 Process Mechanical Cell Crane Room ..................................................................... 111

A5.3.1 External Exposure Data ............................................................................. 111

A5.3.2 Removable Contamination Data................................................................. 111

A5.3.3 Airborne Contamination Data ..................................................................... 111

A5.4 Chemical Process Cell Crane Room......................................................................... 112

A5.4.1 External Exposure Data ............................................................................. 112

A5.4.2 Removable Contamination Data................................................................. 112

A5.4.3 Airborne Contamination Data ..................................................................... 112

A6.0 Chemical Operating Aisle and Chemical Viewing Aisle ........................................................ 112

A6.1 Radiological Conditions ............................................................................................112

A6.1.1 External Exposure Data ............................................................................. 112

A6.1.2 Removable Contamination Data................................................................. 113

A7.0 Chemical Process Cell ......................................................................................................... 113

A7.1 Radiological Conditions ............................................................................................113

A8.0 Control Room....................................................................................................................... 113

A8.1 Radiological Conditions ............................................................................................113

A8.1.1 External Exposure Data ............................................................................. 113
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A9.0 Equipment Decontamination Room and Equipment Decontamination Room 

Viewing Aisle........................................................................................................................ 114

A9.1 Radiological Conditions ............................................................................................114

A9.1.1 External Exposure Data ............................................................................. 114

A9.1.2 Removable Contamination Data ................................................................. 114

A9.1.3 Airborne Contamination Data ..................................................................... 114

A10.0 Extraction and Product Purification Cells.............................................................................. 114

A10.1 Radiological Conditions ............................................................................................115

A10.1.1 External Exposure Data ............................................................................. 115

A10.1.2 Removable Contamination Data................................................................. 115

A10.1.3 Airborne Contamination Data ..................................................................... 116

A11.0 Extraction Chemical Room................................................................................................... 116

A11.1 Radiological Conditions ............................................................................................116

A11.1.1 External Exposure Data ............................................................................. 116

A11.1.2 Removable Contamination Data................................................................. 117

A12.0 Extraction Sample Aisle and Process Sample Cell 2............................................................ 117

A12.1 Radiological Conditions ............................................................................................117

A12.1.1 Removable Contamination Data................................................................. 117

A12.1.2 Airborne Contamination Data ..................................................................... 117

A13.0 Fuel Receiving and Storage................................................................................................. 117

A13.1.1 Radiological Conditions.............................................................................. 118

A13.1.2 Pool Water Concentration Data.................................................................. 119

A13.1.3 External Exposure Data ............................................................................. 120

A13.1.4 Removable Contamination Data................................................................. 121

A13.1.5 Airborne Contamination Data ..................................................................... 122

A14.0 Guard House........................................................................................................................ 122

A14.1 Radiological Conditions ............................................................................................122

A15.0 General Operating Aisle....................................................................................................... 123

A15.1 Radiological Conditions ............................................................................................123

A15.1.1 Exposure Rate Data................................................................................... 123

A15.1.2 Airborne Contamination Data ..................................................................... 123

A16.0 General Purpose Cell........................................................................................................... 123

A16.1 Radiological Conditions ............................................................................................123

A16.1.1 External Exposure Data ............................................................................. 124

A16.1.2 Removable Contamination Data................................................................. 124

A17.0 Head End Ventilation Building and Crane Room .................................................................. 124

A17.1 Radiological Conditions ............................................................................................124

A17.1.1 External Exposure Data ............................................................................. 124

A17.1.2 Removable Contamination Data................................................................. 124

A18.0 Hot Acid Cell ........................................................................................................................ 125

A18.1 Radiological Conditions ............................................................................................125

A18.1.1 External Exposure Data ............................................................................. 125
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A18.1.2 Removable Survey Data ............................................................................ 125

A19.0 Hot Shop.............................................................................................................................. 125

A19.1 Radiological Conditions ............................................................................................125

A19.1.1 External Exposure Data ............................................................................. 125

A19.1.2 Removable Contamination Data................................................................. 125

A20.0 Instrument Shop/Old Instrument Shop ................................................................................. 126

A20.1 Radiological Conditions ............................................................................................126

A20.1.1 External Exposure Data ............................................................................. 126

A20.1.2 Removable Contamination Levels.............................................................. 126

A21.0 Interceptor and New Interceptor........................................................................................... 126

A21.1 Radiological Conditions ............................................................................................127

A21.1.1 External Exposure Data ............................................................................. 127

A22.0 Laundry................................................................................................................................ 127

A22.1 Radiological Conditions ............................................................................................127

A22.1.1 External Exposure Data ............................................................................. 127

A22.1.2 Removable Contamination Data................................................................. 127

A22.1.3 Airborne Contamination Data ..................................................................... 127

A23.0 Load-in Facility..................................................................................................................... 128

A24.0 Low Level Waste Treatment Facility..................................................................................... 128

A24.1 Radiological Conditions ............................................................................................128

A24.1.1 External Exposure Data ............................................................................. 128

A25.0 Low Level Wastewater Treatment Building .......................................................................... 128

A25.1 Radiological Conditions ............................................................................................129

A25.1.1 External Exposure Data ............................................................................. 129

A25.1.2 Removable Contamination Data................................................................. 129

A26.0 Lower Extraction Aisle.......................................................................................................... 129

A26.1 Radiological Conditions ............................................................................................129

A26.1.1 External Exposure Data ............................................................................. 129

A27.0 Lower Warm Aisle................................................................................................................ 129

A27.1 Radiological Conditions ............................................................................................130

A27.1.1 External Exposure Data ............................................................................. 130

A27.1.2 Removable Contamination Data................................................................. 130

A27.1.3 Airborne Contamination Data ..................................................................... 130

A28.0 Main Office Building ............................................................................................................. 131

A28.1 Radiological Conditions ............................................................................................131

A28.1.1 External Exposure Data ............................................................................. 131

A28.1.2 Removable Contamination Data................................................................. 131

A29.0 Maintenance Shop and Warehouse ..................................................................................... 131

A29.1 Radiological Conditions ............................................................................................132
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A30.0 Manipulator Repair Room .................................................................................................... 132

A30.1 Radiological Conditions ............................................................................................132

A30.1.1 External Exposure Data ............................................................................. 132

A30.1.2 Removable Contamination Data................................................................. 132

A31.0 Mechanical Operating Aisle (North, East and West)............................................................. 132

A31.1 Radiological Conditions ............................................................................................133

A31.1.1 External Exposure Data ............................................................................. 133

A31.1.2 Removable Contamination Data................................................................. 133

A31.1.3 Airborne Contamination Data ..................................................................... 134

A32.0 MSM Shop/Contact Size Reduction Facility ......................................................................... 134

A32.1 Radiological Conditions ............................................................................................134

A32.1.1 External Exposure Data ............................................................................. 134

A32.1.2 Removable Contamination Data................................................................. 134

A33.0 Off Gas Aisle and Process Sample Cell 3 ............................................................................ 135

A33.1 Radiological Conditions ............................................................................................135

A33.1.1 External Exposure Data ............................................................................. 135

A33.1.2 Removable Contamination Data................................................................. 135

A33.1.3 Airborne Contamination Data ..................................................................... 136

A34.0 Off Gas Blower Room .......................................................................................................... 136

A34.1 Radiological Conditions ............................................................................................136

A34.1.1 External Exposure Data ............................................................................. 136

A34.1.2 Removable Contamination Data................................................................. 137

A35.0 Off Gas Cell ......................................................................................................................... 137

A35.1 Radiological Conditions ............................................................................................137

A35.1.1 External Exposure Data ............................................................................. 137

A36.0 Permanent Ventilation System Building................................................................................ 138

A36.1 Radiological Conditions ............................................................................................138

A36.1.1 External Exposure Data ............................................................................. 138

A36.1.2 Removable Contamination Data................................................................. 138

A37.0 Process Chemical Room...................................................................................................... 138

A37.1 Radiological Conditions ............................................................................................138

A37.1.1 External Exposure Data ............................................................................. 138

A37.1.2 Removable Contamination Data................................................................. 138

A38.0 Process Mechanical Cell...................................................................................................... 139

A38.1 Radiological Conditions ............................................................................................139

A39.0 Product Packaging and Shipping, Product Packaging and Handling, Uranium 

Product Cell, and Uranium Loadout Areas ........................................................................... 139

A39.1 Radiological Conditions ............................................................................................140

A39.1.1 External Exposure Data ............................................................................. 140

A39.1.2 Removable Contamination Data................................................................. 140

A39.1.3 Airborne Contamination Data ..................................................................... 140
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A40.0 Pulser Equipment Aisle ........................................................................................................ 140

A40.1 Radiological Conditions ............................................................................................141

A40.1.1 Airborne Contamination Data ..................................................................... 141

A41.0 Remote Handled Waste Facility ........................................................................................... 141

A42.0 Sample Storage Cell, Analytical Hot Cells, and Analytical Decontamination 

Aisle..................................................................................................................................... 141

A43.0 Scrap Removal Area............................................................................................................ 141

A43.1 Radiological Conditions ............................................................................................142

A43.1.1 External Exposure Data ............................................................................. 142

A43.1.2 Removable Contamination Data................................................................. 142

A43.1.3 Airborne Contamination Data ..................................................................... 142

A44.0 Solvent Storage Terrace ...................................................................................................... 142

A44.1 Radiological Conditions ............................................................................................142

A44.1.1 External Exposure Data ............................................................................. 142

A44.1.2 Removable Contamination Data................................................................. 143

A45.0 Supernatant Treatment System Support Building................................................................. 143

A45.1 Radiological Conditions ............................................................................................143

A45.1.1 External Exposure Data ............................................................................. 143

A45.1.2 Removable Contamination Data................................................................. 143

A46.0 Upper Extraction Aisle.......................................................................................................... 143

A46.1 Radiological Conditions ............................................................................................143

A46.1.1 External Exposure Data ............................................................................. 143

A46.1.2 Removable Contamination Data................................................................. 144

A47.0 Upper Warm Aisle and Process Sample Cell 1 .................................................................... 144

A47.1 Radiological Conditions ............................................................................................144

A47.1.1 External Exposure Data ............................................................................. 144

A47.1.2 Removable Contamination Data................................................................. 145

A47.1.3 Airborne Contamination Levels .................................................................. 146

A48.0 Utility Room and Stairwells................................................................................................... 146

A48.1 Radiological Conditions ............................................................................................146

A48.1.1 External Exposure Data ............................................................................. 146

A48.1.2 Removable Contamination Data................................................................. 147

A48.1.3 Airborne Concentration Data...................................................................... 148

A49.0 Ventilation Exhaust Cell ....................................................................................................... 148

A49.1 Radiological Conditions ............................................................................................148

A49.1.1 External Exposure Data ............................................................................. 148

A50.0 Ventilation Systems.............................................................................................................. 148

A50.1 Main Ventilation System ...........................................................................................148

A50.2 Head End Ventilation System ...................................................................................150

A51.0 Ventilation Wash Room........................................................................................................ 150
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A51.1 Radiological Conditions ............................................................................................150

A51.1.1 External Exposure Data ............................................................................. 151

A51.1.2 Airborne Contamination Data ..................................................................... 151

A52.0 Waste Tank Farm................................................................................................................. 151

A52.1 Radiological Conditions ............................................................................................151

A52.1.1 External Exposure Data ............................................................................. 151

ATTACHMENT B, LANDAUER EXPOSURE REPORT CODES...................................................... 152
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 11 of 152

LIST OF TABLES

TABLE TITLE

2-1 NFS-defined contamination areas..........................................................................................

PAGE

22

2-2 Respiratory protection thresholds........................................................................................... 22

2-3 Average annual collective dose for maintenance workers for typical tasks, 

1970 to 1971..........................................................................................................................28

2-4 Tasks with the highest collective exposures to maintenance workers, 

1970 to 1971..........................................................................................................................29

2-5 UWA pump changeout data, 1970 and 1971.......................................................................... 30

2-6 LWA pump changeout data, 1970 and 1971 .......................................................................... 30

2-7 Breakdown of the dose to operations workers for typical tasks and duties ............................. 32

2-8 Dose rates and occupancy times for operations workers for various areas 

of the plant, July 1971 ............................................................................................................ 33

2-9 Breakdown for analytical workers, July 1971.......................................................................... 33

2-10 Personnel exposure associated with main ventilation system filter 

changes ................................................................................................................................. 34

2-11 Collective dose for changeouts of HEV system prefilters........................................................ 34

2-12 Collective dose for changeouts of HEV system roughing filters.............................................. 34

2-13 Collective whole-body, skin, and extremity dose for various work groups, 

1967.......................................................................................................................................36

2-14 Collective whole-body, skin, and extremity dose for various work groups, 

1968.......................................................................................................................................36

2-15 Collective whole-body, skin, and extremity dose for various work groups, 

1969.......................................................................................................................................36

2-16 Collective whole-body, skin, and extremity dose for various work groups, 

1970.......................................................................................................................................36

2-17 Collective whole-body, skin, and extremity dose, various work groups for 

1971.......................................................................................................................................37

2-18 Typical exposures per person for various job duties, 1970..................................................... 37

2-19 Typical exposures per person for various job duties, 1971..................................................... 38

2-20 Significant radiation sources in the reprocessing plant as of March 15, 

1972.......................................................................................................................................38

2-21 Exposure rate data for various locations extracted from a health physics 

logbook, July to December of 1973........................................................................................ 39

2-22 Exposure rate data for various locations extracted from a health physics 

logbook, January to October 1974 ......................................................................................... 42

2-23 Exposure rate data for later in the residual radiation era ........................................................ 44

2-24 Estimated quantities of Special Nuclear Material and source material in 

the SDA through 1975............................................................................................................ 45

2-25 Estimated radioactivity content in the SDA through 1975 ....................................................... 45

2-26 Isotopic mix percentages for West Valley plutonium product.................................................. 48

2-27 Radionuclides of concern....................................................................................................... 49

2-28 Dominant radionuclides and conservative estimates of total activity for 

process building facilities and other site facilities.................................................................... 49

3-1 X-ray exam frequencies for West Valley workers ................................................................... 51

4-1 Total particulate releases from West Valley’s main stack 1966 through 

1972.......................................................................................................................................52

4-2 Total 85Kr releases from the West Valley stack 1966 through 1969 ........................................ 54

4-3 Monthly 85Kr release data for 1967 and 1971 ......................................................................... 54
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4-4 Airborne releases of 131I for July through December 1971...................................................... 55

4-5 Particulate discharges for 1970 and 1971 .............................................................................. 56

4-6 External onsite environmental dose for West Valley workers ................................................. 58

5-1 Summary of in vivo bioassay frequencies for West Valley radiation 

workers ..................................................................................................................................64

5-2 Summary of routine in vitro bioassay frequencies for West Valley 

radiation workers....................................................................................................................65

5-3 In vitro analyses and reporting levels for West Valley during the 

operations and residual radiation eras.................................................................................... 66

5-4 Required detection levels for WVDP in vitro analyses as of September 12, 

1990.......................................................................................................................................68

5-5 Required detection levels for priority and routine in vitro samples as of 

August 1, 1997.......................................................................................................................68

5-6 Required detection levels for WVDP in vitro analyses as of March 14, 

1998.......................................................................................................................................68

5-7 WVDP decision levels for positive in vitro samples as of November 23, 

1998.......................................................................................................................................69

5-8 Sensitivity data for April 1967 in vivo counting........................................................................ 70

5-9 Sensitivity data for April 1969 in vivo counts........................................................................... 71

5-10 Nuclide mix information for West Valley MFAP source term as of January 

1, 2000...................................................................................................................................73

5-11 Nuclide mix information for West Valley uranium source term as of 

January 1, 2000 .....................................................................................................................73

5-12 Nuclide mix information for West Valley plutonium source term as of 

January 1, 2000 .....................................................................................................................74

5-13 Nuclide mix information for freshly separated and aged plutonium at West 

Valley.....................................................................................................................................74

6-1 Nuclear-Chicago film badge type codes................................................................................. 80

6-2 Discrepancy codes and explanations ..................................................................................... 80

6-3 Landauer exposure type codes .............................................................................................. 81

6-4 Selection of radiation energies for West Valley facilities......................................................... 83

6-5 Exchange frequencies, MDLs, and maximum annual missed deep dose 

for personnel dosimeters........................................................................................................ 83

6-6 Exchange frequencies, MDLs, and maximum annual missed 

nonpenetrating dose for personnel dosimeters....................................................................... 83

6-7 Uncertainties for positive recorded dose at West Valley......................................................... 84

A-1 Gross beta and gross alpha concentration data for the CUP and SP, 

December 1965 ................................................................................................................... 119

A-2 Radioactivity concentration data for the CUP and SP, January to June 

1966..................................................................................................................................... 119

A-3 Gross beta concentration data for the CUP and SP, September 1965 to 

October 1969 ....................................................................................................................... 120

A-4 Exposure rates over the top of the FRS pools, November 1965 to May 

1969..................................................................................................................................... 120
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 13 of 152

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

ARPR Acid Recovery Pump Room

ASDA Atomic and Space Development Authority (State of New York) 

Bq becquerel

CAA Cell Access Aisle

CCR Chemical Process Cell Crane Room

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

Ci curie

cm centimeter

COA Chemical Operating Aisle

CPC Chemical Process Cell

cpm counts per minute

CUP Cask Unloading Pool

CVA Chemical Viewing Aisle

d day

D&D decontamination and decommissioning

DAC derived air concentration

DOE U.S. Department of Energy

DOELAP U.S. Department of Energy Laboratory Accreditation Program

DOG Dissolver Off Gas

dpm disintegrations per minute

EDR Equipment Decontamination Room

EEOICPA Energy Employees Occupational Illness Compensation Program Act

EMOA East Mechanical Operating Aisle

EPA U.S. Environmental Protection Agency

FRS Fuel Receiving and Storage

ft foot

g gram

gal gallon

GCR General Purpose Cell Crane Room

GOA General Operating Aisle

GPC General Purpose Cell

HEPA high-efficiency particulate air

HEV Head End Ventilation

HLW high-level radioactive waste

hr hour

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

in. inch
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IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

L liter

LLW low-level waste

LLWT Low-Level Waste Treatment (facility)

LLW2 Low-Level Wastewater Treatment Building

LWA Lower Warm Aisle

LXA Lower Extraction Aisle

m meter

MDA minimum detectable activity

MDC minimum detectable concentration

MDL minimum detectable level

MeV megaelectron-volt, 1 million electron-volts

MFAP mixed fission and activation product

MFP mixed fission product

mL milliliter

MOA Mechanical Operating Aisle

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPLB maximum permissible lung burden

mR milliroentgen 

mrad millirad

mrem millirem

MRR Manipulator Repair Room

MSM Master-Slave Manipulator

MTU metric ton of uranium

MWd megawatt-day

nCi nanocurie

NDA NRC-licensed Disposal Area

NFS Nuclear Fuel Services

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Office of Compensation Analysis and Support Claims Tracking System

NRC U.S. Nuclear Regulatory Commission

NYSERDA New York State Energy Research and Development Authority

NYU New York University

OGA Off Gas Aisle

OGBR Off Gas Blower Room

OGC Off Gas Cell

ORAU Oak Ridge Associated Universities

PA posterior-anterior

PCR Process Chemical Room

PEA Pulser Equipment Aisle

PMC Process Mechanical Cell

PMCR Process Mechanical Cell Crane Room
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POC probability of causation

PPC Product Purification Cell

PPH Product Packaging and Handling

PPS Product Packaging and Shipping

PSC Process Sample Cell

PSF Plutonium Storage Facility

PUREX plutonium-uranium extraction

PVS Permanent Ventilation System

R roentgen

RBA Radiological Buffer Area

RHWF Remote Handled Waste Facility

s second

scfm standard cubic feet per minute

SDA State-licensed Disposal Area

SP Storage Pool

SR Scrap Removal

SRDB Ref ID Site Research Database Reference Identification (number)

SSC Sample Storage Cell

SSN Social Security Number

STS Supernatant Treatment System 

Sv sievert

SWP Special Work Permit

TBD technical basis document

THOREX thorium extraction

TLD thermoluminescent dosimeter

TRU transuranic

U.S.C. United States Code

ULO Uranium Loadout Area

UPC Uranium Product Cell

UWA Upper Warm Aisle

UXA Upper Extraction Aisle

VEC Ventilation Exhaust Cell

VOG Vessel Off Gas

VWR Ventilation Wash Room

wk week

WMOA West Mechanical Operating Aisle

WTF Waste Tank Farm

WVDP West Valley Demonstration Project

WVNSC West Valley Nuclear Services Company

XC1 Extraction Cell 1

XC2 Extraction Cell 2 

XC3 Extraction Cell 3

XCR Extraction Chemical Room

XSA Extraction Sample Aisle
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 16 of 152

yr year

μCi microcurie

μg microgram

§ section or sections
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 17 of 152

1.0

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

INTRODUCTION

In this document, the word “facility” is used as a general term for an area, building, or group of 

buildings that served a specific purpose at a site. It does not necessarily connote an “atomic weapons 

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. 

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

premises, grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 

(42 U.S.C. 7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. 

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

[i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all occupationally derived radiation 

exposures at covered facilities in its dose reconstructions for employees at DOE facilities, including 

radiation exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and 

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived:

] guidelines established under 

subsection (c) ...” [42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 18 of 152

1.1 PURPOSE

This site profile provides information about operations at the West Valley reprocessing plant operated 

initially by Nuclear Fuel Services (NFS) and subsequently by DOE under the West Valley 

Demonstration Project. The operations addressed pertain to occupational radiation exposures for 

monitored or unmonitored workers. 

1.2 SCOPE

Section 2.0 provides a description of the site and operations that pertain to possible radiation 

exposures and discusses radiation source terms. Section 3.0 provides guidance for the determination 

of occupational medical dose. Section 4.0 provides guidance for the determination of dose to workers 

outside radiological facilities from external radiation and releases of radioactive materials to the 

environment. Section 5.0 provides guidance for the determination of intakes of radionuclides inside 

facilities. Section 6.0 provides guidance for the determination of external doses from measured doses 

or for periods for which records of measured doses are missing. 

2.0

2.1 SITE HISTORY

SITE DESCRIPTION

In 1962, the State of New York’s Atomic and Space Development Authority (ASDA) and NFS 

partnered to build a privately owned nuclear fuel reprocessing plant. Construction of the plant began 

in June 1963 and took approximately 3 years to complete. The plant was built on 200 acres of land 

leased from the ASDA by NFS. (The State of New York took title to 3,345 acres of land in 1961 and 

established the Western New York Nuclear Service Center.)

NFS was granted a license to receive and store fuel on May 27, 1965. The first fuel for reprocessing 

was received June 3, 1965, and placed in the Storage Pool (SP) on June 5. Approximately 60% of 

the fuel processed at NFS came from the U.S. Atomic Energy Commission (AEC), as part of its 

guarantee to supply fuel until an adequate commercial market grew. A majority of this AEC-supplied 

fuel came from the Hanford N Reactor (DOE 1999). 

An operating license was granted to NFS for the reprocessing plant on April 19, 1966, and 

reprocessing began on April 22. The first three processing campaigns involved fuel from the Hanford 

N Reactor (Clark 1979). 

The West Valley processing plant was designed to be a multipurpose facility with the capability of 

reprocessing nuclear fuel from a number of different fuel cycles (Bailey 1990). The plant primarily 

used the plutonium-uranium extraction (PUREX) process; however, one thorium extraction (THOREX) 

campaign was processed between November 1968 and January 1969. 

From 1966 to 1972, the West Valley facility handled and reprocessed a total of 630 tons of fuel from 

nine different reactors during 28 campaigns. Fuels processed included light-water reactor fuels (from 

both boiling- and pressurized-water reactors), fuels from AEC-owned reactors (such as the Hanford 

N Reactor), and a uranium-thorium fuel cycle core from the Indian Point 1 reactor. Plutonium and 

uranium recovered from irradiated fuels were delivered as nitrates. The recovered uranium was sent 

to the Fernald (Ohio) Plant, and the plutonium was sent to either Hanford or (later) to the ASDA 

Plutonium Storage Facility (PSF). In general, AEC plutonium was sent to Hanford, and utility-owned 

plutonium was retained by the utilities, sold to industry, or sold to NFS for later resale for use in 

plutonium recycle. NFS and the ASDA entered into a plutonium storage agreement on May 3, 1971, 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 19 of 152

in which ASDA would store plutonium at its PSF. The PSF was approximately 2,200 m eastsoutheast of the NFS plant (ASDA and NFS 1971). Storage of plutonium at the PSF was to be 

terminated by December 31, 1974 (NFS 1974a, p. 3.2-1). 

In March 1972, the reprocessing plant was shut down to complete a series of improvements intended 

to increase capacity and meet new regulatory requirements. From 1973 to 1975, 756 spent lightwater reactor fuel assemblies were shipped to the West Valley facility and placed in the SP in 

anticipation of resumption of reprocessing operations (from Bailey 1990). However, after a decision 

by the AEC that a completely new licensing process would be required, and facing more stringent 

requirements on plant effluents, NFS concluded that reprocessing no longer made economic sense. 

In 1977 it transferred management of the facility to the New York State Energy Research and 

Development Authority (NYSERDA). 

In 1980, the U.S. Congress passed the West Valley Demonstration Act, which directed DOE to solidify 

the high-level radioactive waste (HLW) and to decontaminate and decommission the tanks and 

facilities at the West Valley Demonstration Project (WVDP). DOE assumed operational control of the 

West Valley site on February 26, 1982, with West Valley Nuclear Services Company (WVNSC) as its 

contractor (WVNSC 1983). 

One of the first major projects under the WVDP was to address the hundreds of unreprocessed spent 

nuclear fuel assemblies that remained in the SP. During the early 1980s, a campaign was conducted 

to ship spent fuel assemblies back to their point of origin. When this initial shipping campaign was 

halted, 125 fuel assemblies still remained in the SP. Work was subsequently performed in the mid1980s to remove the majority of empty fuel canisters and storage racks in the pool. The SP still 

contained 149 canisters and 11 rows of storage racks after this work was completed. By the early 

spring of 2001, all of the remaining 125 spent nuclear fuel assemblies had been transferred into two 

shipping casks for eventual shipment to the Idaho National Laboratory. By late 2001, the remaining 

149 canisters and storage racks had been removed and decontamination of the SP began [excerpted 

from Jablonski, Al-Daouk, and Moore 2003]. 

Decontamination of the main reprocessing plant took place in the early 1980s to allow for utilization of 

the existing facilities for the HLW vitrification project to the extent practicable (Johnson 1991). 

Decontamination activities took place in a number of former process cells from 1982 through 1987 

(Hoffman 1997). 

Pretreatment of the HLW began in 1988 to partition the cesium and strontium from the other 

constituents and thus create two waste streams: one high level and one low level. This partitioning 

reduced the amount of HLW that required vitrification. This waste pretreatment continued until 1995. 

The effluent (low-level) liquid from the pretreatment process was sent to the Cement Solidification 

System to be solidified in cement. This low-level waste (LLW) was packaged in approximately 20,000 

steel drums, which are being stored at West Valley pending disposal. 

Vitrification of the HLW took place between June 1996 and November 2001. Followed by a program 

to flush residuals through it, the vitrification facility melter was shut down in September 2002. The 

HLW solidification campaign involved the vitrification of approximately 24 million Ci of activity into 

275 glass logs. Each log contains approximately 2,000 kg of glass. The individual logs are encased 

in canisters. Individual canisters had a contact dose rate of approximately 2,600 rem/hr in 2003. 

They are stored in the Chemical Process Cell (CPC) (Petkus et al. 2003). 

In 1999 Vitrification Expended Materials Processing was initiated to begin processing unserviceable 

equipment in the Vitrification Facility. The Remote Handled Waste Facility (RHWF) was subsequently
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 20 of 152

developed to process large-scale, highly contaminated equipment excessed during decontamination 

and decommissioning (D&D) activities. Groundbreaking for the RHWF took place in 2000, as the 

WVDP restructured to focus on D&D. 

In 2003, projects included removal and treatment of sodium-bearing wastewater from underground 

storage tanks and decontamination of the process cell used for purification of the plutonium product 

stream when the reprocessing plant was operating. Debris was removed from the General Purpose 

Cell (GPC) and Process Mechanical Cell (PMC) and packaged, and the Vitrification Cell 

Dismantlement Project was initiated. 

Shipping LLW off the site was a major activity in 2004, with 104,427 ft3 of waste being shipped for 

disposal. Footprint reduction began with the removal of 20 office trailers from the site. 

Decontamination of four of the major process cells continued: the Vitrification Cell, PMC, GPC, and 

Extraction Cell–2. In November 2004, the Vitrification Melter was removed from the Vitrification 

Facility, packaged in a specially designed container, and staged on the site for potential offsite 

disposal. The Vitrification Melter was the last large component removed from the Vitrification Cell. 

The Vitrification Cell Dismantlement project was completed in mid-2005 (DOE 2006). 

LLW shipments for offsite disposal increased significantly in 2005. Site footprint reduction activities 

escalated, and a number of site employees were moved to an offsite building. 

In addition to the reprocessing plant, NFS operated a commercial radioactive waste disposal site at 

West Valley from 1963 to 1975. Disposal of solid material by land burial began in November 1963. 

During this time NFS buried waste from commercial generators totaling 2.4 million ft3

. These wastes, 

which came from nuclear power facilities, hospitals, and academic, industrial, and government 

facilities, and the reprocessing plant itself, were buried in soil trenches. The material ranged in activity 

from low-specific activity material to containers with exposure rates of 8,000 to 10,000 R/hr (Author 

unknown ca. 1968, p. 1). This burial site, known as the State-licensed Disposal Area (SDA), 

encompasses approximately 15 acres. NYSERDA took over maintenance of the SDA in 1983. 

NFS also operated a disposal (burial) area for high-level solid wastes generated during fuel 

reprocessing. These reprocessing wastes, which included fuel cladding, were buried in a 7-acre 

landfill known as the U.S. Nuclear Regulatory Commission (NRC)-licensed Disposal Area (NDA). 

2.2 REPROCESSING PLANT OPERATIONS ERA, 1965 TO 1973

2.2.1

The NFS reprocessing plant was on a 3,345-acre site bounded by a barbed-wire fence and posted as 

a restricted area. The plant itself and its waste storage operations were performed within a 300-acre 

security area at about the center of the site. The security area was defined by an 8-ft security fence. 

Plant and Process Description

The reprocessing plant consisted of a complex of cells with the various supporting and operating 

areas grouped around them (Johnson & Higgins 1972, p. 2). The plant was arranged in the shape of 

a U, with the Fuel Receiving and Storage (FRS) facility on one end and the product removal facilities 

on the other. The mechanical and chemical processing cells were in the middle (Runion 1970). Most 

areas of the process building fall into one of three categories: shielded cells, operating aisles, and 

unshielded rooms (WVDP 2005). The cells consisted of reinforced concrete walls several feet thick. 

The rest of the plant was of cinderblock construction. Chemical operations were directed from the 

Control Room, while mechanical operations were directed from operating aisles adjacent to viewing 

windows in the hot cells (Runion 1970).
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Fuel reprocessing involved shearing and leaching of fuel elements on the front end, followed by 

solvent extraction to remove plutonium and uranium. Fuel arrived at the site in shipping casks. The 

casks were unloaded under water and the fuel was stored in a SP. These operations took place in 

the FRS facility. Fuel bundles would be inverted and passed through a transfer canal into the PMC. 

Shearing and sawing equipment removed hardware and segmented the fuel into fixed lengths. The 

segmented fuel fell through a chute into baskets in the GPC below. The baskets were then passed 

into the CPC where the segmented fuel was dissolved in acid. This process, known as leaching, 

dissolved the fuel, leaving behind the cladding and any structural components. This leftover material, 

referred to as “hulls,” was rinsed and disposed of in the NDA as HLW. The plutonium and uranium in 

the dissolved fuel were separated from the fission products and purified in a series of extraction cells. 

The purified products were then sent to their owners as nitrate solutions. After May 1971, plutonium 

solutions were sometimes also sent to the Plutonium Storage Facility, which was owned and operated 

by the State of New York on the West Valley site. High-level liquid waste generated in the process 

was stored in underground tanks. Some of the acids used in the process were recycled for reuse. 

Thorium-uranium fuel from Indian Point Unit 1 was processed from November 15, 1968, to January 

20, 1969 (Birchler 1970). This is referred to as the “Con. Ed.” campaign in the vernacular of the day. 

Following the THOREX processing, the plant was thoroughly flushed to reduce the amount of highenriched uranium in systems as much as possible. Processing of the thorium fuel required significant 

changes to the process to account for the high levels of thorium, 235U, and 233U present. The fuel also 

contained 239Pu. The high concentrations of fissile species required the use of boron in the dissolving 

process. A number of systems normally used in processing (e.g., wash systems, silica beds) were 

not used due to criticality concerns. High 238Pu activity required resin beds to be regenerated more 

frequently than usual. 

Process Summary:

• Fuel receipt and storage

• Mechanical preparation

• Fuel dissolving 

• Solvent extraction

• Plutonium product concentration, storage, and loadout

• Uranium product concentration, storage, and loadout

As of 1970 the plant had processed fuel with burnups as high as 30,000 MWd/MTU (Runion 1970). 

The last fuel reprocessing campaign at West Valley was completed in November 1971 (Nelson 

1973a, p. 1). The last plutonium scrap recovery operation took place in March 1972 (Nelson 1973b, 

p. 2 of enclosure 2). Afterward, the plant was in a shutdown condition, with operations limited to fuel 

receipt and storage and decontamination activities. Per the AEC, decontamination activities were 

significantly curtailed after May 4, 1973 (Nelson 1973b, p. 2 of enclosure 2). 

2.2.2

Given the extreme radiological environment which existed while the plant was in operation, the 

radiation protection program focused on controlling personnel exposures relative to legal maximums. 

Exposure control was threshold-oriented, i.e., concerned with not exceeding quarterly exposure limits, 

allowable breathing zone air concentrations, maximum permissible body burdens (MPBBs), etc. (e.g., 

see Keely and Wenstrand 1971). 

Radiation Protection Program

NFS defined contamination areas using a system of four “zones” based upon removable 

contamination levels (Table 2-1). Zone 1 referred to equipment and areas with no smearable 

contamination. Zone 2 areas were those where removable contamination levels were such that 
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special access controls were not deemed necessary. Zone 3 areas were those with contamination 

requiring special control, and Zone 4 areas required “stringent” control. Entry points into Zone 4 areas 

had buffer zones classified as Zone 3 areas. Entries into Zone 4 areas were covered under Special 

Work Permits (SWPs). Radiation Work Permits were used otherwise. 

Table 2-1. NFS-defined contamination areas (NFS 1976). 

Category

Removable alpha 

(dpm/100 cm2)

Removable beta 

(dpm/100 cm2) Posting and barriers

Zone 1 ≤10 ≤100 “Clean” areas outside the security fence: not posted.

Zone 2 10–50 100–500 “Clean” areas inside the security fence: not posted.

Zone 3 50–500 500–50,000 Posted, rope barrier, step-off pad.

Zone 4 ≥500 ≥50,000 Posted, rope barrier, step-off pad, entrance was airlock or 

tented.

All repair or maintenance activities in Zone 4 areas or in high radiation areas (exposure rates ≥ 100 

mR/hr) were performed under SWPs. Standard Operating Procedures addressed contamination 

control, doffing, etc. Decontamination was performed following maintenance activities to reduce 

contamination to premaintenance levels. Radioactive wastes generated from maintenance activities 

were packaged and sent to one of the two burial sites for disposal. 

Respiratory protection thresholds were as given in Table 2-2 (NFS 1974b). These values are the 

maximum allowable concentration for the corresponding respiratory protection equipment. 

Table 2-2. Respiratory protection thresholds. 

Radionuclide

No respiratory 

protection (μCi/cm3)

Full face canister 

mask (μCi/cm3)

Supplied air mask 

(constant flow) (μCi/cm3)

Mixed fission products 1E-09 2E-08 1E-06

Pu-239 2E-12 4E-11 2E-09

Uranium 6E-11 1E-09 6E-08

It is of note that smoking was allowed in operating areas of the process building such as the 

Mechanical Operating Aisles (MOAs). A Johnson & Higgins insurance inspection report from 1972 

(Nelson 1972) echoes information seen in other documents that smoking in contaminated areas was 

an issue at West Valley. They recommended more extensive posting and enforcement of “no 

smoking” requirements in surface contamination areas. As of January 1, 1973, smoking, eating, 

drinking, or chewing were prohibited in Zone III or Zone IV areas, but smoking was still allowed in 

designated Zone II areas (NFS 1973a, p. 2). 

2.2.3

Users of this site profile should bear in mind the West Valley reprocessing plant was an extreme 

radiological environment throughout the operations era. Fuel segmentation operations resulted in 

substantial quantities of high specific activity airborne particulate matter, resulting in significant 

operational difficulties associated with the plant ventilation systems and airflow issues. This, coupled 

with other unforeseen circumstances involving radioactivity in systems where it was not anticipated, or 

at unanticipated levels, meant radiological conditions encumbered operation of the facility from the 

outset. Routine, contact maintenance activities had to be performed in high dose rate environments. 

Dose rates in normally occupied areas were also high, and radiological contamination was substantial 

plant-wide from maintenance activities and spills (see Attachment A for details). High backgrounds 

compromised the effectiveness of contamination control measurements [1]. 

Special Considerations for the Operations Era
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At its maximum, the plant employed a permanent work force of less than 200 persons during 

operations. Contract employees had to be brought in once the regular employees reached their 

quarterly dose limits. There are also reports of using persons “not normally assigned to radiation 

work” to complete high exposure tasks (Keely 1967a, p. 1). A 1972 insurance survey report 

remarked, “To facilitate operations and meet AEC established radiation exposure criteria the plant 

utilizes about 1000 temporary laborers each year. Radiation exposure and contamination hazards are 

extensive and ever present in this type of operation” (Johnson & Higgins 1972, p. 1). 

Users of this site profile should recognize the principal focus of the radiation protection program at 

West Valley during its operations era was to minimize the number of individuals receiving exposures 

in excess of legal maximums [2]. This is particularly true for external exposures, which were 

controlled to a quarterly limit of 3 rem to the whole body. Quarterly skin and extremity limits were also 

enforced. Thresholds for internal monitoring were also applied, commensurate with the regulatory 

requirements of that era. Personnel exposures could not be optimized in accordance with normal 

radiation protection practice given the work environment. Therefore, any assumptions regarding the 

likelihood of personnel exposures at West Valley during its operations era based on job title or work 

area should be carefully considered [3]. In particular, 

• For the period 1966 through 1973, users should not assume nonradiation workers or 

individuals whom did not normally work in the process building did not receive appreciable 

occupational exposures without additional justification. 

• Users should be aware that significant contamination or exposure rates (or both) could have 

been present in areas not normally associated with personnel exposures, such as stairways, 

lobbies, etc. Significant contamination could result from large spills of process solutions, in 

which case it would be persistent indefinitely, or from widespread tracking of contamination 

from maintenance activities, which would not be detected until it reached lower background 

areas of the building. This consideration applies primarily to the 1966 – 1973 time period. 

In addition to the following discussions of personnel exposure mechanisms at West Valley during 

operations, users are also referred to Attachment A for detail on the radiological environment in the 

process building and other areas. 

2.2.4

In a March 1972 letter, the AEC criticized NFS for designing its radiation control program vis-à-vis

maximum allowable exposure limits for workers over the course of its operating history rather than 

making efforts to maintain low levels of exposure, consistent with “universally accepted” radiation 

protection practice. In response, NFS acknowledged it appeared that the principal cause of these 

conditions was “performance of plant equipment at less than design expectations” (Miller 1972a, p. 1). 

The overall largest contributors to the chronically high personnel exposures at West Valley during its 

operating years are discussed below. 

Chronic Sources of Personnel Exposures

2.2.4.1 Plant Ventilation Systems

The principal personnel exposure mechanism in the processing plant was “finely divided fuel particles” 

in direct-maintenance work areas adjacent to process cells, in analytical hot cells, in sample 

enclosures, solution storage areas, filter housings, and ventilation ducts. The particles were released 

during various head-end process operations, including shearing of fuel rods and transporting these 

from one process cell to another, dissolving of fuel, and handling the leftover cladding material 

following dissolving. The particles were a problem for both inhalation exposures and extremity 
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exposures due to their high specific activity and significant beta dose rates (Keely and Wenstrand 

1971). 

The buildup of this high specific activity particulate matter in ducts and other parts of the ventilation 

system was a continuing problem. Following treatment in a washer (scrubber), ventilation cell air 

flowed through a 36-in. duct to filtration equipment at the base of the stack. Buildup of activity in that 

duct caused significant radiation levels in the Analytical Lab areas, the Lower Extraction Aisle (LXA), 

and the Control Room. Activity collecting in the washer caused exposure rates of 40 mR/hr in the 

Analytical Aisle, which was directly above the washer, during processing of high exposure fuels (NFS 

1976). Particulate activity in the ventilation system ducting resulted in exposure rates as high as 100 

mR/hr in the LXA (below the duct), 10 to 20 mR/hr in the Analytical Laboratories, and 5 to 10 mR/hr in 

the Control Room (NFS 1976). The plant design was such that periodic filter changes and other 

maintenance activities on ventilation system components were contact maintenance tasks, i.e., they 

could not be performed remotely. Personnel involved in these operations encountered considerable 

exposure rates.

In response to this problem the plant was modified in two ways. First, penetrations were made into 

the main ventilation system to allow periodic flushing to reduce built up particulate activity. These 

duct-flushing operations would temporarily reduce the exposure rates in personnel areas, but would 

also result in high airborne contamination in some areas while flushing was taking place. The second 

modification was to add an auxiliary ventilation system for the head-end process cells. This Head End 

Ventilation (HEV) system went into operation in October 1970. Following startup of the HEV system 

and decontamination of the original ventilation ducts and washer, radiation levels dropped by a factor 

of 5 to 10 in the Analytical Laboratories and in the Control Room (Miller 1972a, p. 5). However, 

radiation levels in the 36-in. duct continued to be higher than desired even after the HEV system 

came on line.

Wenstrand (1972a) states the ventilation duct running between the ventilation washer and the 

Ventilation Exhaust Cell (VEC) continued to be a primary source of radiation on three levels of the 

plant. It says the duct and washer were extensively decontaminated following startup of the HEV 

system to reduce exposure rates in the operating aisles. However, as of March 1972 the radiation 

levels from the duct had doubled since the decontamination. The report points out the duct contained 

a number of 90-degree bends that interfered with airflow. 

2.2.4.2 Acid Recovery System

Due to startup problems with the waste evaporator systems in 1966, radioactive contamination was 

spread throughout the acid recovery system. This unanticipated radioactivity in the recovered acid 

system resulted in significant exposure rates in numerous areas of the processing plant where such 

elevated levels were not expected. 

Carryover of radioactivity from the waste evaporators and buildup of ruthenium on the inner walls of 

the process chemical feed tanks resulted in exposure rates up to 100 mR/hr in the vicinity of the 

tanks. Ruthenium buildup also occurred in the recovered acid storage tanks located in the Hot Acid 

Cell. This cell was constructed of concrete block and did not afford adequate shielding. Exposure 

rates in the vicinity of the Hot Acid Cell therefore ranged from 100 mR/hr to 1,000 mR/hr. Activity in 

the recovered acid storage tanks caused exposure rates up to 10 mR/hr in the Control Room and 

Office areas (NFS 1976). It also caused elevated exposure rates outside the plant, e.g., in front of the 

office building. 
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2.2.4.3 Fuel Receiving and Storage Facility

The FRS facility was a significant source of personnel exposure for workers in the head end of the 

plant. The primary source of these exposures (internal and external) was contamination of the pool 

water from ruptured fuel elements, and failure of the cleanup systems to reduce activity 

concentrations to desirable levels. A significant decontamination effort was undertaken in the FRS for 

both the pool water and surrounding floors after the plant shut down in March 1972. 

2.2.5

This section describes miscellaneous sources resulting in personnel exposures in the reprocessing 

plant. Attachment A to this document provides detailed information on operating areas and 

associated radiological conditions. 

Miscellaneous Sources of Personnel Exposures

Buildup of Contamination in Piping

Radioactivity would accumulate in steam lines in the LXA, Upper Extraction Aisle (UXA), and the 

Chemical Operating Aisle (COA). These steam lines were used to jet process solution. Exposure 

rates from the lines were 10 to 30 mR/hr and contributed to personnel exposures during sampling and 

valving operations (NFS 1976). 

Duct Flushing Operations

Penetrations were made into the main ventilation system in 1967 to provide for periodic flushing of the 

ductwork to reduce exposure rates in working areas. These duct-flushing operations would cause 

elevated airborne radioactivity levels in numerous areas of the plant. In July 1970 the following areas 

showed concentrations that would have resulted in inhalation exposure in excess of 40 maximum 

permissible concentration (MPC)-hours (TKW 1970):

• General Operating Aisle (GOA)

• MOA (east and west sides)

• Process Sample Cells (PSCs) 1, 2, and 3

• Extraction Sample Aisle (XSA)

• Pulser Equipment Aisle (PEA)

• South Stairs

Radioactivity in Cooling Water, Steam and Condensate Piping

Radioactivity in cooling water, steam, and condensate piping were a source of elevated exposure 

rates in the following areas (NFS 1976): 

• COA

• LXA

• MOA

• Off Gas Aisle (OGA)

• UXA

• Utility Room

• High Level Waste areas

• Control Room
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Analytical Activities

Elevated exposure rates existed in the vicinity of sampling stations from activity in sampling bottles 

and sample lines. Exposure rates were also elevated in the analytical laboratory areas from the 

handling and assay of samples (NFS 1976). 

Extraction Cell Support Equipment

The Lower Warm Aisle (LWA), Upper Warm Aisle (UWA), and the PEA contained pumps, valves, and 

lines in shielded areas for the transfer of process solution within the process cells. These areas 

frequently became contaminated from maintenance activities. 

2.2.6

This section contains personnel exposure information associated with specific tasks or job duties at 

the West Valley processing plant in its operations period. 

Personnel Exposures Associated with Specific Tasks or Duties

2.2.6.1 Maintenance Workers

Maintenance employees received exposures associated with repair, maintenance, and replacement 

activities for contaminated plant equipment. These activities often required “personnel exposures 

higher than those normally received by Operations employees” (NFS 1976). 

Analysis of personnel exposure data for maintenance workers for 1970 to 1971 showed the principal 

sources of exposure for these individuals were the crane rooms and the acid recovery system. Most 

of the exposure in the crane rooms was associated with decontamination and shielding of the area 

and equipment before maintenance activities. The total collective dose for prejob decontamination 

and shielding activities in the three crane rooms was 221 person-rem/yr (NFS 1976). 

The maintenance task resulting in the highest personnel exposures was repair or replacement of the 

crane cable reels. Replacing or untangling of crane or hoist cables was also a significant source of 

personnel exposure for maintenance workers. Other tasks associated with significant dose to 

maintenance workers were repair or replacement of crane motors, repair of power manipulators, and 

retrieval of disabled cranes from the process cells. Collective dose for maintenance workers from

decontamination and repair of Master-Slave Manipulators (MSMs) was 10 person-rem/yr. Collective 

dose for maintenance workers in the Equipment Decontamination Room (EDR) was 22 person-rem/yr

(NFS 1976). 

Another major source of personnel exposure for maintenance workers was servicing the steam jets 

used to transfer recovered acid to the acid storage tanks. The jets suffered from high corrosion rates, 

prompting numerous entries into the Acid Recovery Pump Room (ARPR) to change them. This 

activity resulted in collective dose of 106 person-rem/yr for maintenance workers. Another significant 

source of personnel exposure for maintenance workers associated with the acid recovery system was 

replacement of the heat exchangers in the Acid Recovery Cell. The heat exchangers became 

contaminated when seals failed, allowing contamination to enter the steam system. Repair and 

replacement of the heat exchangers resulted in 59 person-rem/yr of collective dose for maintenance 

workers (NFS 1976). 

A cart mounted on a trolley was used for the transfer of samples of process solution between 

analytical cells. Repair and replacement of this cart following derailment was another source of 

personnel exposure for maintenance workers associated with the plants analytical areas. Cell entries 

would be required when the cart would derail, resulting in about 9 person-rem/yr associated with 

prejob decontamination of the cell and subsequent repair of the cart (NFS 1976). 
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Maintenance workers experienced high personnel exposures performing blower replacement activities 

in the Off Gas Blower Room (OGBR) and those associated with the Exhaust Cell Ventilation and HEV

systems. These exposures were the result of the equipment being located in areas of elevated 

exposure rates. The exposure rate in the OGBR ranged from 1,000 mR/hr to 4,000 mR/hr from 

activity in the acid recovery system and the off gas treatment equipment. The VEC and HEV blowers 

were located in areas with a typical exposure rate of 100 mR/hr (from the filter banks) (NFS 1976). 

Another source of personnel exposure for maintenance workers was repair or replacement of pumps 

in the UWA and LWA. The pumps were located in shielded locations adjacent to the process cells, 

and were used to transfer solutions within the cells. 

The following is a chronology of radiation exposure conditions and events involving maintenance 

workers. A number of these involve skin and extremity exposures. 

• October 1965: Maintenance was performed on a pump that had been used to empty a fuel 

cask. The pump read 11 R/hr and the exposure rate to the maintenance workers was a 

maximum of 1.2 R/hr (Loud 1965a). 

• March 1967: “In general, the maintenance work in the Acid Recovery Cell necessitated 

exposures up to 1,000 mrem to people not normally assigned to radiation work

• December 1967: A maintenance worker received an extremity exposure of 44 rad during work 

in the GPC Crane Room (GCR) (Wenstrand 1968a). 

” [emphasis 

added]. “These exposures began toward the end of March” (Keely 1967a, p. 1). 

• December 27, 1967: A maintenance foreman received an indicated 160 rad exposure to his 

hand as shown by a thermoluminescent dosimeter (TLD) worn as a finger dosimeter. Surveys 

of the work area concluded the exposure was from hot particles, as the maximum dose rate in 

the area was found to have been 16 rad/hr. The finger rings of two other individuals 

(maintenance workers) involved with the task showed 2 rad and 3.8 rad (Lewis 1968). 

• June 1968: Installation of a new blower in the OGBR required 26 person-rem of collective 

dose. The typical working area exposure rate was 1 to 2 R/hr (Keely 1968a). 

• August 1968: A maintenance worker received an extremity exposure of 22.5 rem and a skin 

exposure of 6.28 rem while working in the Recovered Acid Storage Area (Keely 1968b). A 

different report says his skin exposure was 5.33 rem, the 22.5 rem extremity dose was to the 

front of his hand, and he received 12.7 rem to the back of his hand. His whole-body exposure 

was 0.95 rem (Urbon 1968a, p. 2). 

• September 1968: A maintenance worker received an extremity exposure of 18.82 rem 

working in the Chemical Cell Crane Room (Keely 1968c). 

• October 1968: Changeout of the Dissolver Off Gas (DOG) blower in the OGBR required the 

use of 60 people and a collective dose of 48 rem whole body (Keely 1968d). 

• December 1968: A maintenance worker exceeded his quarterly extremity exposure limit while 

working in the GCR. He received a dose of 21.750 rem (Keely 1969a). 

• March 1969: A maintenance worker exceeded his quarterly extremity exposure limit when he 

received an exposure of 13 rem in the GCR (Keely 1969b).
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• May 1969: A maintenance worker received a skin exposure of 6520 mrem while changing a 

pump in the hot acid storage area (Keely 1969c, p. 1). 

• June 1969: A maintenance worker received an extremity exposure of 11.7 rem during repair 

work on the GPC crane (Keely 1969d). 

• July 1969: Two persons received extremity exposures of 15 rem and 30 rem while working in 

the GCR (Keely 1969e). 

• October 1971: 61 “supplemental workers” received a collective 58 rem of dose changing a 

pump. NFS staff (operations, maintenance, and health and safety) received an additional 6 

rem (Wenstrand 1971a). 

NFS performed a review of 1970 and 1971 personnel exposure data for personnel involved in the 

“major work categories” of operations and maintenance. The average collective dose for operations 

and maintenance personnel over these 2 years was 985 person-rem and 1,002 person-rem, 

respectively. They assert incidents were minor contributors to these totals, which implies they were 

not biased by single events (NFS 1976). 

Table 2-3 summarizes average annual collective dose for maintenance workers for typical tasks over 

the period 1970 to 1971 (NFS 1976). All collective dose data in this site profile reflect external 

exposure only. 

Table 2-3. Average annual collective dose for maintenance workers for typical tasks, 

1970 to 1971. 

Plant area Typical work performed

Personrem/yr

ARPR Changing recovered acid jets 106

Misc. maintenance: replacing valves, instrument lines, etc. 4

Valving 2

Inspections 1

ARC Repair or replacement of 7C-3 heat exchanger 14

Repair or replacement of 7C-5 heat exchanger 45

Analytical cells Decontamination 5

Repair, maintenance of transfer cart 4

CCR Decontamination and shielding 53

Repair or replacement of crane cable reels 40

Repair or replacement of crane motors 27

Repair PAR 20

Replace or untangle crane cable 8

Repair of misc. items 10

Structural repair of crane 8

Retrieval of disabled crane 2

EDR Decontamination and cleanup 7

Repair of CPC eductors, jumpers, etc. 10

Repair of CPC PAR 5

GCR Decontamination and shielding 106

Repair or replacement of crane cable reels 70

Repair of cell door hoist 32

Retrieval of disabled crane 23

Replace or untangle hoist cables 25

Repair PAR 19
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Plant area Typical work performed

Personrem/yr

Repair or replacement of crane motors 10

Misc. electrical repairs on crane 12

Structural repair of crane 3

LWA Changing pumps or valves 9

Misc. maintenance 3

Valving and inspection 1

OGBR Change blower 6

Change off gas scrubber pumps 5

Misc. maintenance: replacing valves, instrument lines, etc. 5

HEV Replace blowers 3

PMCR Decontamination and shielding 62

Repair or replacement of crane cable reels 40

Replace or untangle hoist cables 31

Repair or replacement of crane motors 13

Misc. electrical repairs on crane 12

PAR repairs 5

Repair PMC cell equipment 2

Structural repair of crane 2

UWA Change pumps 12

Misc. maintenance 3

VEC Maintenance and repair of blowers 7

Misc. maintenance 2

Decontamination 6

VWR Install new filters in washer 21

Replacement and repair of pumps 9

MSM Repair of manipulators 10

Plant Misc. maintenance in plant areas (operating aisles) 62

Total 1,002

Table 2-4 ranks the tasks that resulted in the highest exposures (collectively) to maintenance workers 

in 1970 to 1971 (NFS 1976). 

Table 2-4. Tasks with the highest collective exposures to 

maintenance workers, 1970 to 1971. 

Area Task Dose

ARPR Changing recovered acid jets 106

GCR Decontamination and shielding 106

GCR Repair or replacement of crane cable reels 70

PMCR Decontamination and shielding 62

Plant Misc. maintenance in operating aisles, etc. 62

CCR Decontamination and shielding 53

ARC Repair or replacement of 7C-5 heat exchanger 45

CCR Repair or replacement of crane cable reels 40

PMCR Repair or replacement of crane cable reels 40

GCR Repair of cell door hoist 32

PMCR Replace or untangle hoist cables 31

CCR Repair or replacement of crane motors 27

GCR Replace or untangle hoist cables 25

GCR Retrieval of disabled crane 23

VWR Install new filters in washer 21
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Tables 2-5 and 2-6 show collective dose data for pump changes for the period 1970 through 1971 for 

the UWA and LWA, respectively. 

Table 2-5. UWA pump changeout data, 1970 and 

1971 (Wenstrand 1974, p. 3). 

Date

Exposure 

rate (R/hr)

Collective dose for 

changeout (person-rem)

1/2/1970 2 0.30

3/26/1970 5 0.47

4/7/1970 10 2.46

5/13/1970 4 1.20

7/6/1970 15 1.44

7/20/1970 4 2.70

2/3/1971 5 2.67

3/31/1971 2 1.46

6/26/1971 7 0.91

7/15/1971 3–5 3.26

10/29/1971 5 5.78

Table 2-6. LWA pump changeout data, 1970 and 

1971 (Wenstrand 1974, p. 4). 

Date

Exposure 

rate (R/hr)

Collective dose for 

changeout (person-rem)

3/23/1970 2 0.70

5/12/1970 3 1.20

5/20/1970 15 0.63

5/26/1970 10 0.88

6/16/1970 2 0.75

6/22/1970 2 0.73

6/30/1970 4 0.38

10/22/1970 7 1.81

11/25/1970 2 0.42

11/30/1970 10 2.85

6/7/1971 5 3.18

7/14/1971 3 1.05

8/3/1971 4 0.72

2.2.6.2 Operations

NFS’ review of 1970 and 1971 personnel exposure data found the average collective dose for 

operations personnel over that period was 985 person-rem (NFS 1976). NFS asserted incidents were 

minor contributors to the collective dose for operations workers (as well as maintenance workers). 

Table 2-7 shows the breakdown of collective dose to operations workers for typical tasks and duties

(NFS 1976). 

Table 2-8 shows dose rates and occupancy times for Operations workers for various areas of the 

plant for July 1971 (NFS 1976). Table 2-9 shows a breakdown for analytical workers for the same 

period (NFS 1976). 
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2.2.6.3 Ventilation System Filter Changeouts

A large contributor to personnel exposures within Operations duties was changeout of ventilation 

system filters. These included filters associated with the main ventilation system, the HEV system, 

the DOG system, and the Vessel Off Gas (VOG) system. The average annual collective dose 
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Table 2-7. Breakdown of the dose to operations workers for typical tasks and duties. 

Plant area Typical work performed

Average annual collective dose (person-rem) for operations workers for 1970–1971

General

area

Ventilation

systems

Acid recovery

system

Contact

maintenance areas Total

FRS Fuel receiving, storage, transfer to PMC 250 -- -- -- 250

Labs Sample analysis 50 90 -- 10 150

CR Process surveillance 5 75 -- -- 80

GOA GPC operations 20 60 -- -- 80

XCR Mixing chemicals, valving, readings 12 8 60 -- 80

HEV Filter changes 5 64 -- -- 69

MOA PMC operations 20 5 -- 25 50

Sample stations Taking process samples 20 10 10 -- 40

SR Decontamination of scrap casks and area 30 -- -- -- 30

VEC Change filters 5 23 -- -- 28

-- Miscellaneous 27 -- -- -- 27

OGA Chemical mixing, valving, readings 5 3 8 -- 16

OGBR Change filters -- 15 -- -- 15

-- Waste burial 15 -- -- -- 15

-- Hull burial 10 -- -- -- 10

Offices Supervision -- 1 6 -- 7

COA Valving, readings 4 1 -- -- 5

PPS/PPH Product loadout 5 -- -- -- 5

Yard Interceptor, lagoons, etc. 5 -- -- -- 5

LLWT Liquid waste treatment 4 -- -- -- 4

LXA Valving readings 1 3 -- -- 4

PEA Readings, valving 2 -- 1 -- 3

WTF Readings, valving -- 3 -- -- 3

CVA CPC operations 2 -- -- -- 2

LWA Valving 2 -- -- -- 2

UXA Readings, valving 2 -- -- -- 2

PCR Sampling, valving, readings -- -- 1 -- 1

UR Auxiliary services operation 1 -- -- -- 1

UWA Valving 1 -- -- -- 1

Totals 503 361 86 35 985
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Table 2-8. Dose rates and occupancy times for operations workers for various areas 

of the plant, July 1971.

Area

Occupancy time

(person-hrs./shift)

Dose rate

(mR/hr)

Annual collective

dose (person-rem)

XCR 6 8 53

Control room 18 1.5 30

UXA 2 1 2

PCR -- 100 --

OGA 0.5 30 16

LXA 0.5 3 2

PPH/PPS 5 1 5

Sampling station 6 4 26

COA 3 1 3

WTF Shelter 0.5 5 3

LLLW 8 0.5 4

Interceptor 5 1 5

FRS 12 15 197

GOA 3 20 66

MOA 18 1.5 30

EDRVA 0.5 0.1 --

CVA 8 0.1 1

Office building 7 1 7

Table 2-9. Breakdown for analytical workers, July 1971. 

Area

Occupancy time

(person-hrs./day)

Dose rate

(mR/hr)

Annual collective

dose (person-rem)

QC lab 12 1 5

Hot lab 36 3 39

Shift office 18 4 26

Hot cells 6 4 9

Supervisor's office 3 5 5

Alpha lab 3 3.5 4

Product lab 15 2 11

Emission lab 6 0.7 2

Mass spec. lab 18 0.7 5

Count lab 9 3 10

associated with changing the filters (roughing and absolute) in the VEC was 23 person-rem for 1970

to 1971 (NFS 1976). 

Tables 2-10 through 2-12 below show personnel exposures associated with changeouts of main 

ventilation and HEV system filters during 1970 and 1971. 

The November 1970 exposures are attributed to pre-HEV operation since the activity on the filters 

accumulated before the system was operating. The HEV system came online in late October 1970. 

Additional information pertaining to filter changes follows.

• A memorandum dated December 9, 1966, says the plant ventilation system’s roughing filters 

were changed “contact-wise in a radiation field of 5 to 10 R/hr” (Lewis 1966a).
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Table 2-10. Personnel exposure associated with main 

ventilation system filter changes (NFS 1972a). 

Change date

Collective dose 

(person-rem)

Before HEV operation

03/25/1970 4.1

04/13/1970 1.5

09/15/1970 15.8

11/05/1970 10.1

11/14/1970 4.1

After HEV operation

03/29/1971 4.5

07/28/1971 3.2

11/15/1971 2.2

Table 2-11. Collective dose for changeouts of 

HEV system prefilters (NFS 1972a). 

North train South train

Date PersonremDate Personrem

10/28/1970 0.2 10/29/1970 0.2

11/02/1970 0.2 11/02/1970 0.2

11/25/1970 0.65 11/28/1970 1.98

12/01/1970 1.23 12/05/1970 1.52

12/15/1970 4.25 12/23/1970 6.40

01/02/1970 1.96 01/22/1971 1.01

02/02/1971 1.40 02/24/1971 1.10

04/01/1971 4.70 04/15/1971 0.50

04/30/1971 1.92 05/18/1971 4.02

06/04/1971 0.75 06/22/1971 0.52

07/16/1971 4.17 08/04/1971 4.00

09/17/1971 1.30 09/08/1971 10.00

11/20/1971 1.37 10/07/1971 2.18

-- -- 12/12/1971 4.58

Table 2-12. Collective dose for 

changeouts of HEV system 

roughing filters (NFS 1972a). 

Date Person-rem

12/04/1970 2.29

12/14/1970 1.73

12/22/1970 2.76

03/01/1971 1.87

03/08/1971 2.15

06/21/1971 1.08

07/15/1971 3.20

10/14/1971 5.26

• The VOG filter read 30 R/hr at 2 ft on November 2, 1967. TLD measurements showed 77 R/hr 

and 280 rad/hr at 18 in. (Wenstrand 1967a). 

• The DOG filter read 200 R/hr at 2 in. and 30 R/hr at 2 ft on November 9, 1967 (Wenstrand 

1967a). 
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• A DOG filter changed on 3/9/1968 read 1000 R/hr at 2 in. and 200 R/hr at 4 ft. The filter was 

placed in a lead cask that read 10 R/hr to 30 R/hr at 2 in. By March 25, 1968, the new filter 

read 500 R/hr at 6 in. and 50 R/hr at 4 ft. (Wenstrand 1968b). 

• In May 1968 the typical exposure rate in the filter room of the VEC at the start of a filter 

change was 35 R/hr. Replacement of 30 each roughing and absolute filters required 16 

person-rem of whole-body dose (Keely 1968e). 

• A DOG filter removed on November 27, 1968, read 1000 R/hr at 3 in. (Keely 1968f). 

• Changeout of the main ventilation filters in November 1968 required “a total whole-body

exposure of 3 rem” (Keely 1968f). 

• In January 1972 HEV prefilters were removed from the south plenum using a new method 

utilizing the roof of the Filter Change Room. Personnel dose was 1.2 rem to the whole body. 

The new method is said to virtually eliminate extremity and skin exposure problems

(Wenstrand 1972b). 

2.2.6.4 Overall Workforce

The average

• 1968 = 2.74 rem/person

yearly whole-body exposures for process, operations, and service employees working in 

or frequenting the normal access areas of the processing plant for 1968 through 1971 are as follows. 

• 1969 = 3.81 rem/person

• 1970 = 6.76 rem/person

• 1971 = 7.15 rem/person

These are average whole-body exposures for individual workers. They are not collective dose (Miller 

1972a). The increase in 1970 and 1971 over previous years reflects the processing of higher burnup 

fuels, which resulted in significant increases in personnel exposure throughout the site. 

The average weekly whole-body exposure per person for 1967 for various responsibilities was as 

follows (Wenstrand 1967b): 

• Head End Operations = 115 mR/wk

• Chemical Process Operations = 195 mR/wk

• Maintenance = 110 mR/wk

• Health & Safety = 80 mR/wk

Based on these data, NFS made the following projections for annual exposures per person in 1968 

under the assumption that the weekly exposure date for 1967 did not change. 

• Head End Operations = 5.8 rem/year

• Chemical Process Operations = 9.8 rem/year

• Maintenance = 5.5 rem/year

• Health & Safety = 4.0 rem/year

Table 2-13 through 2-17 below show collective whole-body, skin and extremity dose for various work 

groups for 1969 through 1971. 
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Table 2-13. Collective whole-body, skin, and extremity dose for various work 

groups, 1967 (Wenstrand 1971b, p. 5). 

Dept. No. workers

Collective dose (person-rem)

Whole body Skin Extremity

Analytical 28 66 175 176

Health & Safety 9 28 68 88

Maintenance 26 77 297 748

Production: Chemical 38 181 324 418

Production: Mechanical 29 77 243 418

Overhead + others 79 121 243 352

Total 209 550 1,350 2,200

Table 2-14. Collective whole-body, skin, and extremity dose for various work 

groups, 1968 (Wenstrand 1971b, p. 6). 

Dept. No. workers

Collective dose (person-rem)

Whole body Skin Extremity

Analytical 34 77 130 331

Health & Safety 14 62 116 156

Maintenance 38 135 251 823

Production 107 368 503 1,268

Overhead 71 81 110 243

Total 264 723 1,110 2,821

Benz + non-NFS 106 127 210 579

Table 2-15. Collective whole-body, skin, and extremity dose for various work 

groups, 1969 (Wenstrand 1971b, p. 7). 

Dept. No. workers

Collective dose (person-rem)

Whole body Skin Extremity

Analytical 29 122 164 763

Health & Safety 16 63 129 182

Maintenance 36 148 320 835

Production 74 404 727 1,649

Overhead 70 122 145 323

Total 225 859 1,485 3,752

Benz + non-NFS 82 109 153 251

Table 2-16. Collective whole-body, skin, and extremity dose for various work 

groups, 1970 (Wenstrand 1971b, p. 2). 

Dept. No. workers

Collective dose (person-rem)

Whole body Skin Extremity

Analytical 24 165 173 572

Health & Safety 14 108 230 374

Maintenance 34 204 358 837

Operations 61 532 693 1,459

Overhead 24 46 74 150

Tech. Services 6 43 59 109

Total (NFS) 163 1,098 1,587 3,501

Benz + non-NFS 472 433 1,043 2,630
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Table 2-17. Collective whole-body, skin, and extremity dose for various work 

groups, 1971 (Wenstrand 1972c, p. 2). 

Dept. No. workers

Collective dose (person-rem)

Whole body Skin Extremity

Analytical 25 146 154 544

Health & Safety 15 116 236 361

Maintenance 34 248 517 910

Operations 66 589 900 1,665

Overhead 15 41 57 109

Tech. Services 7 32 47 78

Total (NFS) 162 1,172 1,911 3,667

Benz 991 1,194 2,677 4,690

Table 2-18 shows typical exposures per person for various job duties for 1970. The data are typical 

whole-body, skin, and extremity dose per individual. They are not collective dose. The exposure data 

are for NFS employees. Exposures for contract workers are not included. 

Table 2-18. Typical exposures per person for various job duties, 1970 (Wenstrand 

1971b, p. 3). 

Job classification

Typical yearly exposure (rem)

Whole body Skin Extremities

Analytical Process Control Supervisors 7.3 7.9 26

Process Control Analysts 8.6 9.4 40

Special Lab Supervisors 5.8 6.0 14

Special Lab Analysts 6.3 6.4 15

Health & Safety Supervisors 8.7 12 23

A Technicians 7.9 20 28

B, C Technicians 7.2 16 16

Maintenance Management 8.9 15 52

Supervisors 8.2 15 32

Electricians 6.6 13 27

Instrument Technicians 7.7 11 16

Mechanics 6.9 11 32

Operations Management 8.6 10 24

Supervisors 10.5 15 32

A Operators 8.5 11 21

B, C Operators 7.6 10 21

Table 2-19 shows typical exposures per person for various job duties for 1971. The data are typical 

whole-body, skin, and extremity dose per individual. They are not collective dose. Exposures for 

contract workers are not included.

Table 2-20 shows exposure rate data for the processing plant as of March 15, 1972 (after shutdown). 

From review of the March 1972 report (Wenstrand 1972a) and a follow-up report from July of that year

(Wenstrand 1972d), it appears sometimes the exposure rates given are for the source itself, and other 

times they are general exposure rates in the affected area. Exposure rates that appear to be for a 

specific source are in bold type. These distinctions were drawn by comparing the March 1972 survey 

report with the follow-up report from July 1972.

Table 2-21 shows exposure rate data for various locations extracted from a health physics logbook for 

a period between July and December 1973 (NFS 1974c). More detailed information on exposure 

rates in a given area can be found in Attachment A. 
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Table 2-19. Typical exposures per person for various job duties, 1971. 

Job classification

Typical yearly exposure (rem)

Whole body Skin Extremities

Analytical

Supervisors 9.2 5.6 11.6

Process control 6.8 7.6 31.6

Spec. labs 4.4 4.4 13.2

Health & safety Supervisors 8.0 18.8 34.8

Technicians 7.6 16.0 24.0

Maintenance

Management 8.4 15.2 20.0

Supervisors 7.2 17.2 26.8

Electricians 7.2 18.0 27.2

Instrument technicians 8.0 11.6 13.6

Mechanics 8.0 16.8 34.8

Operations

Management 6.8 8.8 18.0

Supervisors 8.8 16.8 27.2

A operators 9.6 13.2 24.8

B, C operators 8.8 13.2 26.0

Table 2-20. Significant radiation sources in the reprocessing plant as of March 15, 1972. 

Plant level Radiation source

Exposure rate 

(mR/hr)* Affected area(s)

Below grade Contamination around leaking window 20 GOA

HEV Duct 10 GOA

100-ft elevation

FRS 10–20 EMOA

Manipulator Repair Room (MRR) 100 EMOA

HEV Room 10 MSM Shop

Piping in window areas 100 WMOA

Contamination around shear 2.5

Contamination in southwest stairwell 10 Showers and locker room

114.5-ft elevation

Ventilation duct 50 LXA

Contaminated piping and general 

contamination around the C-1 Sampler

10 COA

Contaminated piping in the COA 50 COA

Ventilation washer 5 LXA

131-ft elevation

Ventilation duct 5–10 Laboratory area

Ventilation washer 10 Laboratory area

PMCR Penthouse 5 Laboratory area

Drain in hot laboratory 10

144-ft elevation

Ventilation duct 2.5 Control Room

Hot piping in UXA 20 UXA

Ventilation filter bank 100 UXA

Hot acid cell 10 Process Chemical Room

160-ft elevation

Hot acid tanks, lines, and floor 

contamination

1,000 XCR

Pump pot niche in the PEA 100 XCR

Ventilation duct 50 XCR

Hot acid tanks 1,000 XCR

Waste Tank Farm Ventilation line 500

2.3 REPROCESSING PLANT RESIDUAL RADIATION ERA, 1974 TO FEBRUARY 25, 1982

This era is characterized by miscellaneous decontamination and plant improvement work. The level 

of activity was not as great as the period between shutdown in March 1972 and May 1973. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 39 of 152

Table 2-21. Exposure rate data for various locations extracted from a health physics logbook, July to December 1973 (NFS 1974c). 

Area Date range

Exposure rate 

range: low

Exposure rate 

range: high Remarks

Analytical aisle July 9–December 27, 1973 1–4 mR/hr 2–12 mR/hr

Analytical aisle–north July 11–December 27, 1973 1–5 mR/hr 1–35 mR/hr

Counting room August 3–December 27, 1973 1–2 mR/hr 1–6 mR/hr

Analytical supervisor's office July 11–December 27, 1973 2–10 mR/hr 5–40 mR/hr

Hot lab July 11–December 27, 1973 1 mR/hr 1–8 mR/hr

Emission spec. lab July 11–December 27, 1973 < 1 mR/hr 1 mR/hr

Mass spec. lab July 11–December 27, 1973 < 1 mR/hr 1.5 mR/hr

Product lab July 11–December 27, 1973 1 mR/hr 1–30 mR/hr Maximum at hood lip

QC Lab July 11–December 27, 1973 1 mR/hr 1–13 mR/hr Maximum at south hood

COA July 9–December 31, 1973 1–3 mR/hr 1–10 mR/hr

CVA July 9–December 31, 1973 1–18 mR/hr 1–50 mR/hr

Control room July 9–December 31, 1973 1–3 mR/hr 2–6 mR/hr

EDR viewing aisle July 9–December 31, 1973 < 1 mR/hr < 1 mR/hr

LXA July 9–December 31, 1973 1–20 mR/hr 2–35 mR/hr

FRS July 15–December 27, 1973 2–10 mR/hr 5–60 mR/hr

GOA July 10–December 29, 1973 2–6 mR/hr 2–20 mR/hr

Instrument shop July 10–December 26, 1973 1–2 mR/hr 1–2 mR/hr

Laundry July 10–December 29, 1973 0.5–1 mR/hr 1–3 mR/hr

EMOA July 9–December 27, 1973 1–3 mR/hr 2–10 mR/hr

WMOA July 9–December 27, 1973 1–3 mR/hr 1–15 mR/hr

MSM Shop July 10–December 29, 1973 1–3 mR/hr 2–20 mR/hr

Personnel decontamination

room

July 11–December 27, 1973 < 1 mR/hr < 1 mR/hr

Stairs–east July 10–December 31, 1973 1–2 mR/hr 1–5 mR/hr

Stairs–north July 9–December 27, 1973 1–5 mR/hr 3–15 mR/hr

Ventilation supply room July 10–December 31, 1973 1–5 mR/hr 1–20 mR/hr

UXA July 9–December 31, 1973 1–15 mR/hr 3–30 mR/hr

LLWT building July 11–December 12, 1973 1–5 mR/hr 1–18 mR/hr

WTF shelter July 12–December 31, 1973 5–225 mR/hr 40–500 mR/hr

WTF: over tanks July 12–December 31, 1973 < 1–2 mR/hr 1–3 mR/hr

XCR July 18–December 17, 1973 1–10 mR/hr 1–30 mR/hr

Off gas aisle September 21–October 30, 1973 10–20 mR/hr 1000 mR/hr Maximum was at NW corner; says 

2–80 mR/hr "@ rope"

Pulser equipment aisle September 21, 1973 10–300 mR/hr 10–300 mR/hr Says "west niche" open [only one 

measurement recorded for 1973]

PCR July 11–October, 30 1973 1–5 mR/hr 1–10 mR/hr "Area being cleaned"

PSC-1 July 11, 1973 1–5 mR/hr 1–5 mR/hr
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 40 of 152

Area Date range

Exposure rate 

range: low

Exposure rate 

range: high Remarks

PSC-2 July 11–October 30, 19732–5 mR/hr3–5 mR/hr

PSC-3 July 11–November 3, 19735–45 mR/hr5–50 mR/hr

PPS July 18–October 24, 19731–4 mR/hr1–6 mR/hr

PPH July 18–October 24, 19731–2 mR/hr1–3 mR/hr

RER August 6–October 24, 19731–35 mR/hr1–50 mR/hr

Rooftops September 18–December 26, 1973 < 1–100 mR/hr < 1–500 mR/hr

South stairs ???–October 28, 19731–8 mR/hr2–15 mR/hr First entry not dated.

ULO July 18–October 24, 1973 1 mR/hr1–2 mR/hr

Utility room August 11–October 24, 1973 < 1–2 mR/hr < 1–2 mR/hr

NDA October 24, 1973 0.3–0.5 mR/hr 0.3–0.5 mR/hr Only one measurement recorded for 

1973

XSA July 11–November 3, 19731–6 mR/hr1–10 mR/hr
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 41 of 152

Activities at the plant consisted of minor maintenance, decontamination projects, and laboratory 

analyses, including some for NFS-Erwin (Nelson 1976, p. 6 of inspection report). As of May 1975 

they still received three to five fuel shipments per week, usually by truck in NFS shipping casks. 

Railroad shipments were rarely made by this time. There was approximately 115 tons of fuel in the 

SP, which was about 45% full (Johnson & Higgins 1975; Nelson 1976, p. 1). 

A May 12, 1975, Nuclear Insurance Survey report by Johnson & Higgins states the highest level of 

personnel exposure at West Valley was around 400 mrem/quarter received by the fuel unloading 

operators, and that maintenance workers received about 200 mrem/quarter (Johnson & Higgins 

1975). 

The DOG system went inoperative in November 1975. The VOG system remained in use to ventilate 

gases from the vessels and tanks in the process building (WVNSC 1994).

Fuel receipts had ceased by 1980, though the plant was still storing spent fuel (Duckworth 1980). The 

fuel remained in storage until the early 1980s when it was removed by the DOE. 

Table 2-22 shows exposure rate data for various locations extracted from a health physics logbook for 

a period between January and October 1974 (NFS 1974c). 

Table 2-23 shows exposure rate data for later in the residual radiation era. The data are cited by 

Golden, Frank, and Prina (1982). 

2.4 REPROCESSING PLANT DOE ERA, FEBRUARY 26, 1982, TO PRESENT

DOE took over operational control of the processing plant site on February 26, 1982 (WVNSC 1983). 

This included the high-level burial area [4]. Responsibility for the LLW burial area transferred to the 

State of New York in 1983 (NYSERDA 2004). 

Decontamination of the main reprocessing plant took place in the early 1980’s to allow for utilization of 

the existing facilities for the HLW vitrification project to the extent practical. Decontamination and fuel 

handling activities in the early years of the WVDP resulted in higher personnel doses than 

experienced later in the project (Hoffman 1997).

As of 1988, the following areas of the main plant remained contaminated and represented sources of 

potential personnel intakes (Johnson 1991): 

• CPC

• PMC

• GPC

• Extraction Cell 1 (XC1)

• Extraction Cell 2 (XC2)

• Hot Acid Cell

• Plutonium Product Handling

• Solvent Storage Terrace

Inspections performed in 1986 found that much equipment in the head end cells (GPC and PMC) had 

fallen into disrepair. Cranes were not reliable, crane room doors were inoperable, and power 

manipulators and MSMs were inoperable or inaccessible (Vance 1986). Exposure rate surveys inside 

the cells found general levels ranged from 40 R/hr to 340 R/hr at a height of 6 ft. Hot spots were 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 42 of 152

Table 2-22. Exposure rate data for various locations extracted from a health physics logbook, January to October 1974 (NFS 

1974c). 

Area Date range

Exposure rate 

range: low

Exposure rate 

range: high Remarks

Analytical aisle January 4–October 20, 1974 1–7 mR/hr 2–10 mR/hr

Analytical aisle–north January 4–October 20, 1974 1 mR/hr 1–15 mR/hr From 6/2/1974 forward, 20 mR/hr "in 

paint on PMC door"

Counting room January 4–October 20, 1974 1 mR/hr 1–4 mR/hr

Analytical supervisor's office January 4–October 20, 1974 5–20 mR/hr 5–40 mR/hr

Hot lab January 4–October 20, 1974 1 mR/hr 1–4 mR/hr

Emission spec. Lab January 4–October 20, 1974 1 mR/hr 1 mR/hr

Mass spec. Lab January 4–October 20, 1974 < 1 mR/hr 1 mR/hr

Product lab January 4–October 20, 1974 1 mR/hr 1–8 mR/hr

QC Lab January 4–October 20, 1974 1 mR/hr 1–2 mR/hr

COA January 10–October 18, 1974 1–3 mR/hr 1–5 mR/hr

CVA January 10–October 18, 1974 1–15 mR/hr 1–60 mR/hr October 18, 1974 entry says 1–4 

mR/hr, but this is an outlier with 

respect to the other entries, which are 

reasonably consistent.

Control room January 10–October 18, 1974 1–3 mR/hr 1–6 mR/hr

EDR viewing aisle January 10–October 18, 1974 < 1 mR/hr < 1 mR/hr

LXA January 10–October 18, 1974 1–35 mR/hr 1–65 mR/hr

FRS January 3–October 18, 1974 5–20 mR/hr 5–35 mR/hr

GOA January 3–October 18, 1974 2–7 mR/hr 2–20 mR/hr

Instrument shop January 17–October 2, 1974 1 mR/hr 1–2 mR/hr

Laundry January 2–October 18, 1974 < 1 mR/hr 1 mR/hr

EMOA January 3–October 18, 1974 1–4 mR/hr 1–10 mR/hr

WMOA January 3–October 18, 1974 1–2 mR/hr 1–5 mR/hr

MSM shop January 3–October 18, 1974 2–8 mR/hr 2–11 mR/hr

Personnel decontamination

room

January 2–October 20, 1974 < 1 mR/hr 1 mR/hr

Stairs–east January 10–October 18, 1974 < 1–3 mR/hr 1–4 mR/hr

Stairs–north January 3–October 18, 1974 1–4 mR/hr 1–8 mR/hr

Ventilation supply room January 10–October 18, 1974 1–12 mR/hr 1–20 mR/hr

UXA January 10–October 18, 1974 1–20 mR/hr 1–30 mR/hr

UWA May 10–October 18, 1974 1–15 mR/hr 1–20 mR/hr

LLWT building January 13–October 31, 1974 < 1 mR/hr < 1–5 mR/hr

WTF shelter January 13–October 31, 1974 5–200 mR/hr 30–300 mR/hr

WTF: over tanks January 13–October 31, 1974 < 1–3 mR/hr 1–5 mR/hr

XCR January 13–October 20, 1974 1–5 mR/hr 1–15 mR/hr
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 43 of 152

Area Date range

Exposure rate 

range: low

Exposure rate 

range: high Remarks

Off Gas aisle January 15–July 30, 1974 10 mR/hr 1000 mR/hr Maximum was at NW corner.

Pulser equipment aisle January 15–October 1, 1974 2–20 mR/hr 2–25 mR/hr

PCR January 15–July 30, 1974 1–3 mR/hr 1–5 mR/hr

PSC-1 January 15–July 30, 1974 1–5 mR/hr 4–8 mR/hr

PSC-2 January 15–July 30, 1974 3–10 mR/hr 5–15 mR/hr

PSC-3 January 15–July 30, 1974 1–3 mR/hr 5–20 mR/hr

PPS January 11–July 30, 1974 1 mR/hr 1–5 mR/hr

PPH January 11–July 30, 1974 < 1 mR/hr 1–2 mR/hr

RER January 11–July 30, 1974 10–20 mR/hr 2–60 mR/hr

Rooftops May 14–August 21, 1974 < 1–100 mR/hr < 1–100 mR/hr

South stairs May 14–July 30, 1974 1–3 mR/hr 2–5 mR/hr

ULO January 11–July 30, 1974 < 1 mR/hr 1 mR/hr

Utility room February 12–July 30, 1974 1–2 mR/hr 1–3 mR/hr

NDA February 12–September 30, 1974 0.3–0.5 mR/hr 0.5–1.5 mR/hr

XSA January 15–July 30, 1974 1–5 mR/hr 1–8 mR/hr

Old instrument shop May 20–July 30, 1974 20–75 mR/hr 50–100 mR/hr

VEC June 3–July 30, 1974 5–250 mR/hr 5–250 mR/hr Start date may be June 18.

Southwest stairs July 30, 1974 2–45 mR/hr Only one measurement recorded for 

1974.
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 44 of 152

Table 2-23. Exposure rate data for later in the residual radiation era. 

Area Exposure rate (mR/hr) Date

Extraction chemical room 5 Ca. 1981

Extraction cell 1 11,300 October 1978

Extraction cell 2 90 April 1978

Extraction cell 3 12 May 1978

Product purification cell 100 Not given

Acid recovery pump room 500 to 1000 March 1976

Liquid waste cell 150 to 1000 March 1976

Chemical process cell 12,000 to 32,000 Ca. 1981

Process mechanical cell 600,000 (maximum) November 6, 1978

General purpose cell 650,000 (maximum) November 6, 1978

Miscellaneous areas 2 to 10 May 22, 1980

Acid recovery cell 100 to 150 February 1974

Off gas aisle 1 to 2 September 1980

Upper warm aisle 12 to 90 May 1981

Lower warm aisle 5 to 40 April 16, 1974

Chemical operating aisle 10 to 20; 300 maximum June 1981

Lower extraction aisle 10 September 18, 1972

Upper extraction aisle 2.5 to 60 May 1981

Mechanical operating aisle 0.1 to 10 May 1981

General purpose cell crane room 50 to 75 March 1978

Pulser equipment aisle 2 to 10 May 22, 1980

Southwest stairway 500 to 1000 Not given

Equipment decontamination room 10 to 100 1977

Scrap removal 1 to 10 1980

found as high as 2,050 R/hr. Solids and liquids (in sump areas) were sampled, showing high levels of 

137Cs and 60Co, plus fissile and transuranic (TRU) species (Vance 1986). 

The canisters of vitrified waste are stored in the CPC. An individual canister reads around 2,300 

rem/hr at contact (Petkus et al. 2003). The logs of vitrified material read around 5,000 R/hr before

being loaded in the canisters. High backgrounds are still a confounding factor with respect to 

radiation contamination surveys, though not to the extent they were when the plant was operating [5]. 

2.5 STATE-LICENSED (COMMERCIAL) WASTE DISPOSAL AREA, 1963 TO 1983

The SDA occupies approximately 15 acres of the Western New York Nuclear Service Center 

immediately adjacent to the 175 acres DOE is using for the WVDP. NFS opened the SDA in 1963 

and continued to operate it until 1975, when disposal operations were terminated as a result of 

problems managing water that infiltrated the disposal trenches. Approximately 2.4 million ft3 of LLW 

were disposed of in the SDA. This included material such as dry active waste from the NFS 

reprocessing plant. NYSERDA currently has responsibility for the monitoring and maintenance of the 

SDA (NYSERDA 2004). 

During operations, packaged waste was placed in long trenches excavated in the soil that were then 

backfilled. Water from rain and snowmelt was able to infiltrate the trenches through cracks and 

discontinuities in the trench covers and upper levels of the soil. Because of the highly impermeable 

nature of the soil surrounding the trenches, water accumulated in them and became contaminated 

from contact with the wastes. In early 1975, this accumulation, coupled with the site operator's (NFS) 

inability to gain regulatory approval to remove, treat, and discharge the trench water on a controlled 

basis to an adjacent stream, led to uncontrolled seepage of contaminated water from some of the 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 45 of 152

trenches. Waste burial operations promptly ceased and the SDA has remained shut down ever since

(NYSERDA 2004). 

Between 1975 and 1981, NFS pumped water out of the trenches several times and treated it before

release to the adjacent stream. Redesigning and reworking of the covers reduced, but did not 

eliminate, water accumulation in the trenches. Water management problems continued to be a 

concern at the SDA. In 1983, NYSERDA assumed management responsibility for the SDA from NFS

(NYSERDA 2004). 

In June 1973 the New York State Department of Environmental Conservation published a study of the 

waste inventory buried in the SDA and the NDA (Kelleher and Michael 1973). It said the greatest 

volume of waste came from waste disposal firms, most of which was associated with hospital, 

institutional and industrial activities. The use of the site by nuclear power plants was less than 10% of 

the total volume, though the authors noted the usage by nuclear power was increasing. 238Pu was 

asserted to be the nuclide of greatest concern in terms of its quantity and toxicity. The radionuclide 

present in the greatest quantity in terms of activity was tritium.

The SDA was closed March 8, 1975, upon discovery of tritium and 90Sr in surface waters from 

leakage from the SDA. Studies of water pumped from the trenches in 1973 showed positive 

indications of gross beta and gross alpha activity (Duckworth 1976). A February 1974 report on 

radionuclide analyses on water from SDA trenches showed the following radionuclides: tritium, 90Sr, 

129I, 125I, 137Cs, 134Cs, 106Ru, 54Mn, 57Co, 60Co, 22Na, 75Se (NFS 1974d). 

Tables 2-24 and 2-25 show estimated quantities of radioactive materials buried in the SDA through 

1975. The authors did not decay-correct their estimates, which date back to 1963. The data are 

highly uncertain, as they were based upon assumptions of nuclide mix, waste volume ratios, etc. 

They are presented to give users an idea of the isotopes present in the SDA trenches more so than 

quantities or isotopic ratios. It is not clear what is meant by “Misc.” or why they lumped together 125I 

with 226 Ra and 241Am (Duckworth 1978a, p. 1). 

Table 2-24. Estimated quantities of Special Nuclear Material and source material in the SDA through 

1975 (Greco 1978).

Special nuclear material Source material

Pu-238 Pu-239 U-235 U-233 Th-232 U-238 + Unat

Grams Ci Grams Ci Grams Ci Grams Ci lbs Ci lbs Ci

2,136 37,339 2,018 123.906 51,202 0.103 255 2.422 15,393 0.769 1,010,594 151.590

Table 2-25. Estimated radioactivity content in the SDA through 1975 (Duckworth 1978a, p. 2).

Tritium C-14 Co-60 Cs-137 Sr-90

I-125, Ra-226

Am-241 Ru-106 Misc. Pu-238

171,000 1,200 227,000 12,000 29,000 500 3,000 232,000 35,000

A review of solid waste disposal at the West Valley site through 1977 was generated as part of a 

Safety Analysis Report update in 1978 (Duckworth 1978b, p. 1). It showed a total of 2,349,000 ft3 of 

waste totaling 710,600 Ci of activity had been buried in the SDA from 1963 through 1975. This 

included 314,400 ft3 of waste from the reprocessing plant itself. A “special burial” of reactor parts took 

place at the SDA in 1973, involving an estimated 330,000 Ci of activity. 

Radiological Data for the SDA

Three containers of HLW were received in September 1965 having a maximum exposure rate of 45 

R/hr. The maximum exposure rate for the waste handlers was 200 mR/hr (Loud 1965b, p. 1).
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 46 of 152

Two containers of HLW were received in October 1965 showing a maximum exposure rate of 11 R/hr. 

The maximum exposure rate to waste handlers was 500 mR/hr. Twelve containers shipped as LLW

were received during the month, showing exposure rates from 220 mR/hr to 750 mR/hr (Loud 1965a). 

Four shipments of HLW were received in November 1965 with the maximum exposure rate from the 

containers being 10 R/hr. The maximum personnel exposure rate was 1.5 R/hr to individuals rigging 

one of the containers (Loud 1965c). 

Seven shipments of HLW were received for disposal in December 1965. The highest exposure rate 

of those containers was 70 R/hr. The maximum personnel exposure rate was 3 R/hr to individuals 

performing rigging for one of the containers. On December 10, 1965, the working area exposure rate 

for a crane operator moving HLW from storage into a trench was 25 mR/hr. The exposure rate for 

workers involved in removing HLW containers from transportation casks on December 31, 1965, was 

a maximum of 250 mR/hr. The exposure rate from the containers was a maximum of 5 R/hr (Loud 

1966a). 

High-level waste received for disposal in February 1966 showed exposure rates up to 170 R/hr. 

Personnel exposure rates for individuals working with these containers were up to 800 mR/hr. The 

fence around the HLW “compound” had to be relocated due to increasing exposure rates. The 

exposure rate through the concrete wall around the compound reached 190 mR/hr. The exposure 

rate at the barbed wire fence was less than 100 mR/hr, and at the chain link (outer fence) it reached 1 

mR/hr (Loud 1966b, p. 1). 

A container of HLW (received for disposal) in March 1966 read 240 R/hr, which exceeded NFS’ 

license limit of 200 R/hr. They received permission to accept the container from the New York State 

Department of Labor (Loud 1966c). 

2.6 HIGH-LEVEL WASTE DISPOSAL AREA, 1966 TO PRESENT

During the AEC/NRC-licensed era, material was buried at the NDA in holes (shafts) instead of 

trenches. Holes were dug 3 ft wide by 7 ft long by 50 ft deep. Waste containers were placed in the 

holes up to 6 to 8 ft below the ground level. They were then backfilled to a thickness of 4 ft above the 

original grade. Additional cover was added if necessary until a reading made with a GM survey 

instrument could not be distinguished from background. The holes were placed at least 6 ft apart. 

Equipment that was too large to fit in a standard-size hole was buried ad hoc. The principal material 

buried at the NDA was cladding hulls. However, a wide variety of other materials and items were also 

disposed of in the NDA, including ruptured irradiated fuel assemblies (Kelleher and Michael 1973, p. 

8).

The review of solid waste disposal at the West Valley site through 1977 showed a total of 139,500 ft3

of material was buried at the NDA, totaling 551,300 Ci of activity (Duckworth 1978b, p. 1). Of this 

material, 6,400 ft3 was hulls (pieces of cladding left over after the fuel was leached out), 130,200 ft3

was miscellaneous items such as filters and equipment, and 2,900 ft3 was designated as “Under 200 

mR” (Duckworth 1978b, p. 3). 

The WVDP built a Liquid Pre-Treatment System on the northeast corner of the NDA. The purpose of 

the system is to remove organics (solvent) from groundwater around the NDA. As of 2004 the system 

had never been used (WVDP 2005). 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 47 of 152

Radiological Data for the NDA

As of early 1968 hulls and scrap from the head end cells of the process plant were transferred for 

burial at least once a week (Runion 1968, p. 8 of Action Plan - Part 2). The burial area routinely 

exceeded 100 mR/hr in the area occupied by the remote crane operator, who was located in a 

shielded cubicle 100 ft from the crane. Additional shielding was supposed to have been installed by 

June 1, 1968, to ensure that the crane operator was not exposed to exposure rates in excess of 300 

mR/hr. As of July 24, 1968, a lead glass window had been installed at the NDA to reduce the 

exposure to the crane operator. The shielding was said to have reduced the exposure rate from the 

hulls to “about 2 mR/hr,” which is said to be indistinguishable from background (Clark 1968a, p. 1). 

In July 1969 a dissolver removed from the CPC read 30 R/hr at contact with its shield cask and 50 

mR/hr at 50 ft. Fluid leaked from the dissolver while transferring it to the burial area, contaminating 

the road surface to 50 rad/hr. Cleanup involved scrubbing and removal of some of the surface (Keely 

1969e). 

In November 1971 a spill contaminated several acres of ground north of the HLW burial area. Levels 

of 1 mR/hr to 3 mR/hr required extensive cleanup (Wenstrand 1971c, p. 2). 

On June 26, 1972, cement blocks were removed from the OGA for burial. They were wrapped in 

plastic and then brought through the roof of the OGA where they were transferred from the roof to the 

“Red Stake Truck.” The blocks read 5 R/hr (general area), and were described as grossly 

contaminated by NFS staff, having been used to shield an acid recovery line in the OGA. Contact 

dose rate measurements showed 15 R/hr, 40 R/hr, and 250 R/hr for whole body, skin of whole body, 

and extremities, respectively. The self-reading dosimeter of the worker whom unloaded the blocks 

showed 100 mR for the unloading operation. Afterward, the individual was found to be contaminated, 

and a nasal swab showed 180 dpm alpha and 170,000 dpm beta. Chemical analysis of the nasal 

swab showed 19.6 nCi of 106Ru, 0.24 nCi of 90Sr, 0.109 nCi of 239Pu, and 544 nCi of 95Zr/Nb (Nelson 

1973a, p. 14 of attachment). Estimates from whole-body counts were an exposure of 71 MPC-hr “of 

combined concentrations” of 239Pu, 106 Ru, and 95Zr/Nb. 

On July 6, 1972, a waste burial worker became contaminated while oiling a lift crane, but the 

contamination was not detected until the following afternoon. Surveys of his home found 

contamination on his bedding and jacket. A towel read 25,000 cpm with a GM survey instrument. A 

nasal swab collected on July 7 showed 5,800 dpm. Two urinalysis samples were collected on July 7 

showing 11 dpm/24 hr and 14 dpm/24 hr “of fission products, 134,137Cs.” A chest count performed July 

10 showed 32 nCi of 137Cs. 

2.7 ISOTOPIC INFORMATION FOR PLANT AREAS

2.7.1

Little radionuclide-specific information is available for surface or airborne contamination encountered 

in the processing plant prior to the characterization activities conducted under the WVDP. When the 

plant was operating, alpha contamination was assumed to be 239 Pu and beta was treated as 90 Sr for 

radiation protection purposes (Clark 1968b). In general, contamination within the process building 

may be characterized as being associated with product streams, feed streams, or waste streams [6]. 

Operations and Residual Radiation Eras, 1966 to February 25, 1982

Available nuclide mix information for product streams show that the plutonium produced at West 

Valley was in general approximately 75% 239 Pu by weight (see Table 2-26). In terms of alpha activity, 

the data show that, in general, the dominant nuclide was 238Pu (NFS 1969) (HNS 1971). This is 

consistent with the reactor grade plutonium being separated. An exception is the mix asserted for the 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 48 of 152

plutonium involved in a puncture wound incident in January 1972. This material was predominantly 

239 Pu in terms of both mass and alpha activity (Wenstrand 1972e). 

Table 2-26. Isotopic mix data for West Valley plutonium product. 

Isotope

May 30, 1969 1971 January 24, 1972a

Weight 

fraction (%)

Alpha activity 

fraction (%)

Weight 

fraction (%)

Alpha activity 

fraction (%)

Weight 

fraction (%)

Alpha activity 

fraction (%)

Pu-238 0.7 59.9 0.6 58.1 0.03 6.0

Pu-239 76.8 24.1 76.1 24.7 90.7 69.9

Pu-240 13.7 15.9 14.1 17.1 8.4 24.1

Pu-241 7.7 0.1 7.9 0.1 0.8 0.02

Pu-242 1.1 0.02 1.2 0.02 0.1 0.00 a

From a sampling needle responsible for a puncture wound

Plutonium associated with head end operating areas (those upstream of the extraction cells) would

likely have been in a metal or oxide form (absorption type S) [7]. Plutonium associated with the 

process stream, such as would have been encountered in the process, analytical, and product 

handling areas, would have been in a nitrate form (absorption type M) unless it was known to have 

been aged contamination [8]. This freshly separated plutonium would have had very little 241 Am [9]. 

If present, 241Am should be considered as the same absorption type as the plutonium matrix in which 

it has ingrown. In other words americium should be treated as absorption type S if it is associated 

with type S plutonium. 

If the chemical form of plutonium is unknown then either type M or S should be selected based on 

which type maximizes dose to the organ of concern. Bioassay data should be used to determine 

absorption type in cases where sufficient data exist to establish excretion or retention. Because the 

plutonium isotopic mixture encountered at West Valley was predominantly 238Pu on an activity basis, 

the Super S absorption type should not be applied (ORAUT 2007a). This approach should be applied 

for West Valley plutonium cases over the entire time period of concern. See also section 5.5.

Alpha contamination from process areas upstream of the extraction cells (in terms of process flow) 

would have included other alpha species representative of irradiated nuclear fuel. These too would 

have been in metal or oxide form [10]. 

As of December 1966 approximately 11,000 kg of commercial reactor fuel had been processed. The 

first 1,190 kg of uranium produced from this fuel contained 24 ppb plutonium, which exceeded NFS’ 

specification. NFS planned to blend this material with other uranium for which the plutonium content 

was within specifications to get an acceptable mixture. In addition, all of the plutonium product 

recovered the week ending December 16, 1966, exceeded specifications for fission product 

contamination (Lewis 1966b). 

From November 15, 1968, through January 20, 1969, the processing plant ran a THOREX fuel cycle 

to reprocess thorium-uranium fuel from Indian Point Unit 1. Thus, during this time thorium, thorium 

decay products, and 233U would have been encountered in the processing plant in addition to the 

other nuclides normally encountered. There was also higher than usual amounts of 235U and 238Pu 

present during this campaign (Birchler 1970). 

Beta contamination at West Valley would have been a mixture of fission and activation products 

representative of irradiated nuclear fuel. The nuclides present would have included tritium, 58Co, 60Co, 

90Sr/Y, 95Zr/Nb, 99Tc, 106Ru/Rh, 125Sb, 134Cs, 137Cs, and 141,144Ce [11]. In 1966, the principal gammaemitters in the FRS pool water were found to be 58Co, 60Co, and 137Cs (Loud 1966b). In May 1972 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 49 of 152

isotopic measurements of FRS pool water included concentration data for 134Cs, 137Cs, 106Ru/Rh, 

144Ce/Pr and 125Sb (Jaroszeski 1972). Isotopes of ruthenium, rhodium, zirconium, and niobium were

large contributors to the radioactivity in the acid recovery system (NFS 1976). 99Tc was prominent in 

the Uranium Product Cell (UPC) (WVDP 2005). 

Keely and Wenstrand (1971) state “entrainment” (resuspension) factors for radioactive particles found 

in “direct-maintenance” work areas in the plant ranged from 5E-06 m-1 to 1E-03 m-1 depending on 

work activity and airflow. The work areas they refer to included those adjacent to process cells, in 

analytical hot cells, in sample enclosures, solution storage areas, filter housings, and ventilation ducts. 

Resuspension factors were used in conjunction with smear surveys to predict work area air 

concentrations. Users should assume a particle size of 5 μm activity median aerodynamic diameter 

throughout the processing plant unless site-specific information is identified. 

2.7.2

One element of the WVDP was extensive characterization of plant areas and waste products. 

Characterization of HLW performed in 1986 determined the following radionuclides of concern 

(Johnson 1991).

DOE Era, February 26, 1982, to present

Table 2-27. Radionuclides of concern. 

Radionuclide Chemical form

Am-241 All forms

Co-60 Co(NO3)2

Cs-137 All forms

Pu-238 PuO2

Pu-239/240 PuO2

Sr-90 SrSO4

Th-228 Th(NO3)4

Th-232 Th(NO3)4

U-234 UO2(OH)2

U-235 UO2(OH)2

U-238 UO2(OH)2

Strontium titanate was not present in the process building, so absorption type S should not be applied. 

Assessments for radiation protection purposes determined that the major beta-gamma nuclides were 

90Sr/Y, 137Cs, 99Tc, and tritium. Alpha emitters of concern were isotopes of uranium and plutonium, 

and other major actinide series elements (241Am, 232Th, et al.) and their decay products (Hoffman 

1997).

Table 2-28 lists dominant radionuclides and conservative estimates of total activity for process 

building facilities and other site facilities. Some of the activity estimates give two values: one before

deactivation and one after. All activity estimates are decay-corrected to September 30, 2004. It is 

implied the dominant radionuclides are listed in order of decreasing activity, though this is not 

explicitly stated in the reference (WVDP 2005). 

Table 2-28. Dominant radionuclides and conservative estimates of total activity for process building 

facilities and other site facilities. 

Facility or area Dominant radionuclides

Estimated total activity 

as of 9/30/2004

Acid recovery cell Pu-241, Am-241, Pu-239, Cs-137, Pu-240 192 Ci before deactivation

66 Ci after deactivation
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 50 of 152

Facility or area Dominant radionuclides

Estimated total activity 

as of 9/30/2004

ARPR Pu-241, Pu-238, Am-241, Cs-137, Pu-239 31 Ci

Analytical decontamination aisle Cs-137, Sr-90, Pu-241, Am-241, Pu-239 0.1 Ci

Analytical hot cells Cs-137, Sr-90, Pu-241, Am-241, Pu-239 39 Ci

CPC crane room Sr-90, Cs-137, Pu-241, Am-241, Pu-239 5.5 Ci

COA Cs-137, Sr-90, Pu-241, Am-241 Not significant

CPC Cs-137, Sr-90, Pu-241, Am-241, Pu-238 116 Ci

Contact size reduction facility 

(former MSM repair shop)

Cs-137, Sr-90 Not significant

Control room Cs-137, Sr-90 Not significant

EDR Sr-90, Cs-137, Pu-241, Am-241, Pu-239 39 Ci

XC1 Sr-90, Cs-137, Pu-241, Am-241, Pu-239 4,900 Ci before deactivation

71 Ci after deactivation

XC2 Pu-241, Sr-90, Am-241, Cs-137, Pu-238 2.0 Ci

XC3 Sr-90, Cs-137, Pu-241, Pu-238, Am-241 1,274 Ci before deactivation

71 Ci after deactivation

FRS Cs-137, Sr-90, Pu-241, Am-241, Pu-239 282 Ci before deactivation

192 Ci after deactivation

GPC Cs-137, Sr-90, Pu-241, Am-241, Pu-238 6,000 Ci after deactivation

GCR and crane room extension Cs-137, Sr-90, Pu-241, Am-241, Pu-239 7.1 Ci

GOA Cs-137, Sr-90, Pu-241, Am-241, Pu-239 Not significant

HEV building and HEV crane room Sr-90, Cs-137, Pu-241, Am-241, Pu-238 610 Ci (assuming fully-loaded filters)

Hot acid cell Pu-241, Am-241, Pu-238, Pu-239, Pu-240 17 Ci (in tanks)

Liquid waste cell Sr-90, Pu-241, Cs-137, Am-241, Pu-238 1,014 Ci, mostly in tanks 

280 Ci after tanks are removed.

LWA Pu-241, Am-241, Pu-238, Pu-239, Pu-240 85 Ci

MOA Cs-137, Sr-90, Pu-241, Am-241 Not significant

Miniature cell Cs-137, Sr-90, Pu-241, Am-241, Pu-239 8.5 Ci

OGA Pu-241, Pu-238, Am-241, Cs-137, Sr-90 40 Ci

OGBR Pu-241, Sr-90, Am-241, Cs-137, Pu-239 75 Ci

Off gas cell (OGC) Pu-241, Pu-238, Am-241, Cs-137, Sr-90 459 Ci

PMC Cs-137, Sr-90, Pu-241, Am-241, Pu-238 2,100 Ci

PMCR and MRR Cs-137, Sr-90, Pu-241, Am-241, Pu-238 5.2 Ci

Product purification cell (PPC) Pu-241, Am-241, Pu-239, Pu-240, Pu-238 43 Ci

SR Room Cs-137, Sr-90, Pu-241, Am-241, Pu-238 0.1 Ci

Southwest stairs Pu-241, Pu-238, Am-241, Cs-137, Pu-239 5.2 Ci

UWA (and pump niches) Pu-241, Am-241, Pu-238, Sr-90, Pu-239 26 Ci

Uranium loadout area (ULO) Sr-90, Cs-137, Pu-241, Am-241, Pu-238 0.1 Ci (in tanks and pumps)

UPC Cs-137, Sr-90, Tc-99, Pu-238, Am-241 12 Ci

VEC Cs-137, Sr-90, Pu-241, Am-241, Pu-238 66 Ci

Ventilation wash room (VWR Cs-137, Sr-90, Pu-241, Am-241, Pu-238 74 Ci

Low level waste treatment (LLWT) 

building

Cs-137, Sr-90 None available

Permanent ventilation system 

(PVS) building

Cs-137, Sr-90 None available

Supernatant treatment system 

(STS) support building

Cs-137, Sr-90, Am-241, Pu-241, Cm-244 

(98% of the activity is Cs-137 and Sr-90)

213 Ci (in the valve aisle)

3.0

Presently, little information has been garnered on medical X-rays required as a condition of 

employment for West Valley workers. NFS (1976) indicated all radiation workers received 

preemployment physicals and annual physicals thereafter, with chest X-rays taken every 2 years. 

Review of claim files in the NIOSH-Office of Compensation Analysis and Support Claims Tracking 

System (NOCTS) shows workers under DOE received preplacement examinations, which included 

chest X-rays, and that workers included in respiratory protection programs (e.g., asbestos workers) 

OCCUPATIONAL MEDICAL DOSE
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 51 of 152

received routine follow-up exposures. Chest X-rays for West Valley workers have always been 

performed at an offsite location. 

Some of the claim files for West Valley workers make reference to “two view” posterior-anterior (PA) 

chest X-rays, [e.g., the worker received a “two view chest x-ray (posterior/anterior)].” It is believed the 

term “two view” reflects a misinterpretation of the stated PA orientation. There is no indication a 

second view (i.e., a lateral exposure) was ever taken in routine physical examinations. 

Based on the above, users should apply the following assumptions when assigning occupational 

medical dose for West Valley workers. These assumptions should not be applied if specific 

information for a worker is available.

• For the period 1965 (onset of reprocessing plant operations) through February 1982, assume 

workers received a single PA chest X-ray at preemployment, every 2 years thereafter, and at 

termination. 

• For March 1982 forward, assume workers received preemployment and termination 

examinations that included a single PA chest X-ray. If the worker was included in a respiratory 

protection program, assume he received additional PA chest X-rays every 2 years if under age 

45 and annually if 45 or older. 

The examination frequencies given for workers in respiratory protection programs are based on 

practices used at other DOE sites. For example, for 1990 through 2002, asbestos workers at Oak 

Ridge National Laboratory received chest X-rays every 3 years if they were under age 40, every 2 

years if they were between 40 and 49, and annually thereafter. As of 2002, the frequency was 

changed to annual examinations for asbestos workers 45 and older (ORAUT 2006a). An assumption 

of an examination every 2 years if under 45 and annually thereafter is therefore believed to be 

favorable to the claimant.

X-ray examination frequencies are summarized in Table 3-1. 

Table 3-1. X-ray examination frequencies for West Valley workers. 

Period Technique

Examination frequency

All workers

Additional examinations for workers in respiratory protection 

programs

Age <45 Age >45

1965–February 1982 PA chest

Preemployment.

Every 2 years.

Termination.

None None

March 1982–present PA chest Preemployment.

Termination. Every 2 years Annually

Medical X-ray dose should be assigned using the organ dose values given in Table 6-5 of ORAUTOTIB-0006 (ORAUT 2005a) for the appropriate time period in conjunction with the examination

frequencies given in Table 3-1. Information in the claim file would get precedence in cases where 

examination frequencies could be shown to differ from those assumed in Table 3-1. 

4.0

This section describes internal and external radiation exposures to workers while outdoors on the 

West Valley reprocessing plant or WVDP site. “Site” in this context means within the security fence. 

It does not refer to the Western New York Nuclear Service Center property within which the plant site 

OCCUPATIONAL ENVIRONMENTAL DOSE
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 52 of 152

is located. NFS considered the boundary of the state property its site boundary during operations, not 

the area defined by its security fence [12]. 

4.1 INHALATION AND SUBMERSION DOSE

Airborne releases of radioactive material at West Valley were discharged to the atmosphere from the 

plant’s 200-ft main stack and a number of ground-level release points associated with specific 

facilities or operations. 

4.1.1

The following discussion focuses on airborne emissions from the West Valley stack during operations. 

Stack releases after the plant ceased operating were much smaller in magnitude. The airborne 

effluents during operations are assessed using maximizing assumptions of 2,600 hr/yr of occupancy, 

a breathing rate of 1.5 m3

/hr, and a particle size of 1 μm activity median aerodynamic diameter 

(AMAD). The latter leads to maximized organ dose for the nuclides considered. These maximizing 

assumptions would not normally be used in a site profile. In most cases a best estimate approach 

would be provided utilizing 2,000 hr/yr of occupancy and a particle size of 5 μm AMAD. A maximizing 

approach is used here to emphasize the magnitude of West Valley’s airborne effluents vis-à-vis dose 

to personnel who spent time outdoors while onsite.

Stack Releases

NFS’ technical specification limits for airborne effluents were as follows (Runion 1970, Table 1): 

• A continuous release limit of 0.1 μCi/s for particulates (alpha plus beta)

• A daily release limit of 12,600 Ci/d for 85Kr

• An annual release limit for 131I of 3.3 Ci

The daily limit for 85Kr reflects the batchwise nature of the fuel dissolving process. 

NFS performed both continuous monitoring and sampling of its stack effluent. Monitoring data were 

usually provided in monthly plant status reports. Results were typically reported as a fraction 

(percentage) of the applicable technical specification limit. Occasionally, total activity would be 

included in monthly effluent statements, but in general, total activity was only reported on an annual 

basis. 

Particulate releases reported prior to March 1969 represent 1 day of decay for short-lived activity. 

Beginning in March 1969, particulate samples were decayed for 3 days prior to counting (Keely 

1969b). 

4.1.1.1 Particulates

Table 4-1 shows total particulate releases (alpha plus beta) from the main stack for 1966 through 

1972. Fuel dissolving was carried out between 1966 and 1971. The processing plant shut down in 

1972 after completion of plutonium recovery operations. 

Table 4-1. Total particulate releases from 

West Valley’s main stack 1966 through 1972. 

Year Total particulates released (Ci)

1966 0.15

1967 0.45

1968 1.14
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1969 0.12

1970 0.19

1971 0.01

1972 0.15

Table 4-1 shows 1968 was the maximum year for particulate releases from the West Valley 

processing plant, despite higher burnup fuels being processed in later years. The relatively large 

particulate release of 1.14 Ci for 1968 was the result of numerous filter failures (see North and Clark 

1968 and NFS 1968). 

NFS reported total beta and alpha particulate releases of 1.86E-1 Ci and 1.16E-3 Ci, respectively, for 

1970 (Miller 1972b). These values give a beta-to-alpha ratio of 160. For 1971, NFS reported total 

beta and alpha particulate releases of 1.39E-2 Ci and 1.92E-4 Ci, respectively (Miller 1972b). These 

values give a ratio of 72. NFS (1976) states the ratio of 72 for 1971 was “also typical of other years.”

Applying the beta-to-alpha ratio of 72 to the particulate release for 1968 gives 1.12 Ci beta and 1.56E2 Ci alpha. These were the maximum particulate releases from the West Valley stack over its history. 

In terms of annualized release rate, the values are 3.55E-8 Ci/s beta and 4.95E-10 Ci/s alpha. 

A calculation of site boundary alpha and beta particulate concentrations performed ca. 1974 (Author 

unknown ca. 1974) states the minimum dispersion (i.e., maximum concentration) is given by a 

dispersion coefficient of 2.5E-7 s/m3

. This is a larger value than seen in more recent, albeit more 

formal, references. For example, the 1985 environmental monitoring report for the WVDP reports a 

maximum dispersion coefficient of 1.4E-7 s/m3 (WVNSC 1986, Table 4-1). 

Since the annual average breathing zone concentration on the site property will be significantly less 

than that where the maximum dispersion coefficient occurs, using the maximum coefficient to 

estimate onsite concentration is favorable to the claimant. In this case, the maximum dispersion 

coefficient is 2.5E-7 s/m3

. 

Calculation of maximized annual intake from the particulate releases for 1968 is shown below. The 

occupancy of 2,600 hr/yr reflects 50 hr/wk for 52 wk/yr. 

Type

Total 

release (Ci)

Annual 

release rate 

(Ci/s)

Dispersion 

coefficient 

(s/m3)

Concentration 

(Ci/m3)

Concentration 

(Bq/m3)

Breathing 

rate (m3/hr)

Occupancy 

(hr/yr)

Annual intake 

(Bq)

Beta 1.12 3.55E-08 2.5E-07 8.9E-15 3.3E-04 1.5 2,600 1.3E+00 

Alpha 1.56E-02 4.95E-10 2.5E-07 1.2E-16 4.6E-06 1.5 2,600 1.8E-02

The magnitude of the maximum particulate intakes from stack emissions are such that selection of 

isotopic mix is largely immaterial. If one assumes Type M 239Pu for 100% of the alpha intake, the 

resulting effective dose is 8.5E-2 mrem for a particle size of 1 μm AMAD. The maximum organ dose 

is 2.7 mrem to the bone surface. Likewise, if one assumes Type S 106Ru for 100% of the beta intake, 

the resulting effective dose is 7.4E-3 mrem and the maximum organ dose is 3.4E-2 mrem to the lung. 

These calculations reflect effective and organ dose coefficients from International Commission on 

Radiological Protection (ICRP) Publication 68 (ICRP 1995) and ICRP Publication 72 (ICRP 1996), 

respectively. 

The calculations above show the magnitude of the particulate releases from the main stack, and 

associated personnel dose were of minimal consequence even under unrealistic, maximizing 

assumptions. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 54 of 152

4.1.1.2 Krypton-85

Reports of the total activity of 85K released for a given year have only been identified for 1966 through 

1969. These are given in Table 4-2. The largest release for these years was 328,000 Ci in 1967. 

Table 4-3 shows 85Kr release data for 1967 and 1971 in terms of fraction of the daily release limit for 

each month. These are the only data available for 1971. No data have been found for 1970, but the 

1971 releases should be comparable. There is little correlation between particulate and noble gas 

emissions, so particulate data are not a reliable means for estimating 85Kr releases.

Table 4-2. Total 85Kr releases from the 

West Valley stack 1966 through 1969. 

Year Total Kr-85 released (Ci)

1966 77,000

1967 328,000

1968 193,000

1969 300,000

Table 4-3. Monthly 85Kr release data for 1967 and 1971. 

Month

Kr-85 releases (% of daily limit) Kr-85 releases (Ci/month)

1967 1971 1967 1971

January 7.7% 50.0% 30,076 195,300

February 6.0% 52.0% 21,168 183,456

March 0.0% 10.3% 0 40,232

April 5.3% 22.0% 20,034 83,160

May 7.4% No fuel dissolved 28,904 0

June 7.2% 30.0% 27,216 113,400

July 16.4% 40.0% 64,058 156,240

August 17.9% 42.0% 69,917 164,052

September 0.0% 7.3% 0 27,594

October 4.9% 41.0% 19,139 160,146

November 5.0% 27.0% 18,900 102,060

December 7.2% N/A 28,123 0

Totals 327,537 1,225,640

The curies per month data given in Table 4-3 were computed by multiplying the reported fraction by 

12,600 Ci (the daily release limit) and then by the number of days in the respective month. The 

Ci/month values for 1967 sum to 327,537 Ci, matching the reported value of 328,000 Ci in Table 4-2. 

The total 85Kr released in 1971 corresponds to an annualized release rate of 0.04 Ci/s. Applying the 

maximum dispersion coefficient of 2.5E-7 s/m3 (see Section 4.1.1.1) gives a maximum concentration 

on the site property of 1.0E-8 Ci/m3

, or 370 Bq/m3. 

The ICRP Publication 68 effective dose rate factor for 85Kr is 2.2E-11 Sv/d per Bq/m3 (ICRP 1995). 

Applying this to the maximum 85Kr concentration gives an effective dose rate of 8.14E-9 Sv/d, or 

8.14E-4 mrem/d. Applying a maximizing assumption for occupancy of 2,600 hr/yr gives an effective 

dose of 0.09 mrem. 

4.1.1.3 Iodine-131

The fuel processed at West Valley had decayed for a minimum of 12 months prior to dissolving. 

Longer cooling times were typical, particularly for fuel with high burnup (i.e., that from commercial 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 55 of 152

power reactors). Currently, the data identified for airborne releases of radioiodines from West Valley 

during operations are incomplete. Data are available for 1966 through 1969 and for part of 1971 (July 

through December). The partial data for 1971 are shown in Table 4-4.

The data in Table 4-4 suggest the 131I released from the West Valley stack was a small fraction of the 

technical specification limit of 3.3 Ci/yr. Further, the iodine releases data given in monthly summary 

reports for 1966 through 1969 are never more than a small fraction of the technical specification limit. 

These values are reported as “less than” some fraction of the limit, though the fractions vary from 

month to month. The largest value observed was less than 7% of the limit reported for August and 

Table 4-4. Airborne releases of 131I 

for July through December 1971. 

Month Curies discharged

July 1.66E-04

August 5.62E-04

September 7.80E-05

October 1.54E-04

November 2.37E-04

December 7.80E-05

September 1968. However, 7% of 3.3 Ci is significantly larger than the releases given in Table 4-4 for 

1971. 

Given the variable and incomplete nature of the reported airborne iodine releases from West Valley, a 

maximizing evaluation was performed using the technical specification limit of 3.3 Ci/yr. In terms of 

release rate, 3.3 Ci/yr equates to 1.0E-7 Ci/s. Calculation of the associated maximized annual intake 

is shown below. 

Type

Total 

release (Ci)

Annual 

release rate 

(Ci/s) 

Dispersion 

coefficient 

(s/m3

) 

Concentration 

(Ci/m3

) 

Concentration 

(Bq/m3) 

Breathing 

rate (m3/hr)

Occupancy 

(hr/yr)

Annual intake 

(Bq)

I-131 3.3 1.05E-07 2.5E-07 2.6E-14 9.7E-04 1.5 2,600 3.8E+00

The ICRP Publication 68 effective dose coefficient for 131I in vapor form is 2.0E-8 Sv/Bq (ICRP 1995). 

Using this value, an intake of 3.8 Bq gives an effective dose of 7.6E-8 Sv, or 7.6E-3 mrem. The 

Publication 68 organ dose coefficient for the thyroid for 131I in vapor form is 3.9E-7 Sv/Bq. This yields 

a maximum organ dose of 1.5E-1 mrem to the thyroid. 

4.1.1.4 Tritium

NFS did not include tritium in its stack monitoring data during the operations era. Tritium was 

released from fuel assemblies during segmentation operations in the head end process cells. The 

U.S. Environmental Protection Agency (EPA) Office of Radiation Programs took measurements of 

tritium releases from West Valley over four processing campaigns from June through November 1971 

(EPA 1972). They correlated their tritium release data to fuel burnup and then used NFS’ processing 

records to estimate tritium releases over the past 6 years. They estimated a total of 2,800 Ci had 

been released, for an average of 460 Ci/yr. They determined the speciation was 76% tritiated water 

vapor (HTO) and 24% gaseous.

For a maximizing assessment, it is assumed the entire 2,800 Ci of tritium is released in a single year 

and that it is 100% HTO. Calculation of the associated maximized intake is shown below.

 
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Type

Total 

release (Ci)

Annual 

release rate 

(Ci/s)

Dispersion 

coefficient 

(s/m3)

Concentration 

(Ci/m3)

Concentration 

(Bq/m3)

Breathing 

rate (m3/hr)

Occupancy 

(hr/yr)

Annual intake 

(Bq)

Tritium 2,800 8.88E-05 2.5E-07 2.2E-11 8.2E-01 1.5 2,600 3.2E+03

Applying the ICRP Publication 68 effective dose coefficient for HTO (1.8E-11 Sv/Bq; ICRP 1995) 

gives an effective dose of 5.8E-8 Sv, or 5.8E-3 mrem. Multiplying by 1.5 to account for skin 

absorption gives 8.7E-3 mrem. 

4.1.2

There were several sources of airborne effluents at or near ground level at West Valley over its 

history. These included:

Ground Level Releases

• Discharges from laundry facilities

• The FRS vent

• The Extraction Chemical Room (XCR) vent

• The XCR tank vent

• The Process Chemical Room (PCR) tank vent

• The Low Level Waste Treatment (LLWT) Facility

The above were sources of ground level airborne effluents while the plant was in operation. There 

were other sources of ground level releases during the WVDP, but these releases were much smaller 

in magnitude than those during operations. The following discussion focuses on the operations era. 

Table 4-5 summarizes particulate discharges from the stack and ground level release points at West 

Valley in 1970 and 1971 (Miller 1972b). 

Table 4-5. Particulate discharges for 1970 and 1971. 

Source

Total discharges for 1970 Total discharges for 1971

Gross alpha (Ci) Gross beta (Ci) Gross alpha (Ci) Gross beta (Ci)

Stack 11.6E-4 18.6E-2 1.92E-4 1.39E-2

Laundry 6.1E-6 2.4E-5 9E-7 8.9E-5

LLWT -- -- 1E-7 1.7E-5

FRS vent 2.5E-6 1.7E-4 9.2E-6 2.5E-4

XCR vent 6.7E-6 1.7E-4 6.9E-5 6.6E-4

PCR tank vent 5E-8 3E-6 5E-8 3E-6

XCR tank vent 2.8E-6 8.5E-2 2.8E-6 8.5E-2

Totals 1.178E-3 2.714E-1 2.74E-4 9.99E-2

Table 4-5 shows the largest source of beta particulate emission other than the stack was the XCR 

tank vent, which released 8.5E-2 Ci of activity for both 1970 and 1971. The XCR vent was the largest 

source of alpha particulate, with 6.9E-5 Ci being released in 1971. In terms of annualized release 

rate, these values are 2.7E-9 Ci/s and 2.2E-12 Ci/s for beta and alpha, respectively. 

The 1985 environmental monitoring report for the WVDP reports a maximum dispersion coefficient for 

a ground level release of 6.44E-6 s/m3

. This value corresponds to a receptor azimuth of 337.5° at a 

distance of 805 m (WVNSC 1986, Table 4-2). Use of this value with the above release rates to 

calculate intake is shown below.
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 57 of 152

Type

Total 

release (Ci)

Annual 

release rate 

(Ci/s)

Dispersion 

coefficient 

(s/m3)

Concentration 

(Ci/m3)

Concentration 

(Bq/m3)

Breathing 

rate (m3/hr)

Occupancy 

(hr/yr)

Annual intake 

(Bq)

Beta 8.5E-02 2.70E-09 6.44E-06 1.7E-14 6.4E-04 1.5 2,600 2.5E+00

Alpha 6.9E-05 2.19E-12 6.44E-06 1.4E-17 5.2E-07 1.5 2,600 2.0E-03

Treating the beta intake as 100% Type S 106Ru of 1 μm AMAD particle size gives an effective dose of 

1.6E-7 Sv, or 1.6E-2 mrem. Treating the alpha as Type M 239Pu with a particle size of 1 μm AMAD 

gives an effective dose of 9.4E-8 Sv, or 9.4E-3 mrem. 

4.2 AMBIENT DOSE

Nuclear-Chicago’s “Explanation of Exposure Report Data” on the reverse of its exposure reports 

states ordinarily they considered any exposure on a control badge to be from a source other than that 

which caused exposure to a monitored individual. Control readings were therefore subtracted from 

readings of personnel badges (Author unknown ca. 1967). Likewise, the “Explanations and Remarks” 

on the reverse of Landauer’s exposure reports says control readings were normally subtracted from 

personnel badge readings. However, it adds personnel badge readings may be normalized to 

Landauer’s own control films (only) if it appeared the controls which accompanied the personnel 

badges had been “exposed differently than the personnel packet” (See Item 6 of Attachment B). 

Exposure rates outside the process building during the operations era were significant with respect to 

exposure to control dosimeters stored in the Guard House. The exposure rate at the Guard House 

during operations was nominally 2 mR/hr from radioactivity in the plant’s acid recovery system. 

Heacker (1972) discusses the problem of dose accumulating on the control badges stored in the 

Guard House. He says the control badge for the week of December 20 to 26, 1971, showed 60 mrem 

because there was a drum reading 30 R/hr on a truck parked overnight at the cooling tower. The 

control badge reading for the following week was 20 mrem (Heacker 1972). 

Per Heacker (1972), it is estimated control badges received a dose of 20 mrem/wk from elevated 

background during operations. Prorating this for 50 hours/wk of occupancy yields 6 mrem/wk which 

needs to be added to dosimeter and missed dose for monitored workers at West Valley over the 

period 1966 through 1973.

It is assumed the exposure rate at the storage location for the control badges was reduced following 

initial cleanup activities completed in 1973. An estimate of the dose rate to control badges in the 

Guard House for the period 1974 until 1982 can be made from later area monitoring data.

After DOE took over site operations in February 1982, dosimetry was only issued to persons whom 

entered radiologically controlled areas. An area monitoring program served to monitor exposures in 

uncontrolled areas to demonstrate compliance with applicable limits. The program used TLDs 

stationed in numerous (indoor) locations about the plant site. Reported results were net values above 

background stated in terms of mrem/500 hr of occupancy. Control TLDs were stored in a shielded 

location (e.g., see WVNSC 1999). 

The earliest period for which area monitoring data are available for the Guard House is 1983 to 1985. 

The annual doses measured for that location were 14 mrem, 11 mrem, and 14 mrem for 1983 to

1985, respectively (Roberts 1986). These values are based on 2,000 hours of occupancy. Adjusting 

for 2,600 hours of occupancy, the maximum annual value of 14 mrem becomes 18 mrem. This dose 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 58 of 152

should be added to dosimeter and missed dose for monitored West Valley workers for the period 

1974 to 1981. 

Personnel dosimetry was required for entry into the plant security area during the operations and 

residual radiation eras. Users should not need to assign external dose for an unmonitored worker 

during this time. For the DOE era, monitoring data representative of what an unmonitored worker 

would have received from ambient exposure while outdoors on the West Valley site have not been 

identified. To be favorable to the claimant, it is assumed an unmonitored worker was exposed to an 

annual dose equal to the applicable administrative limit for the period in question. For 1982 through 

1993 the administrative limit for annual dose to an unmonitored individual is assumed to have been 

170 mrem (see Roberts 1986). For 1994 forward the annual limit was 100 mrem per 10 CFR Part 

835. 

4.3 SUMMARY OF EXPOSURES FROM ENVIRONMENTAL SOURCES

The maximizing calculations described in Section 4.1 show intakes associated with airborne effluents 

at West Valley were negligible. 

Monitored workers should be assigned an additional 312 mrem/year to dosimeter or missed dose for 

the period April 1966 through 1973. For 1974 to 1981, 18 mrem/year should be added. The added 

dose should reflect the monitoring period, i.e., 6 mrem/week or 26 mrem/month for April 1966 through 

1973, and 0.35 mrem/week or 1.5 mrem/month for 1974 through 1981. The dose should be prorated 

as appropriate for partial years of employment. The additional dose should be treated as photons in 

the 30- to 250-keV energy group (ORAUT 2006b). 

Personnel dosimetry was required for entry into the plant security area between 1966 and February 

1982, when DOE took over site operations [13]. Users should not need to assign external dose for an 

unmonitored worker during this time. For 1982 to 1993 170 mrem/yr should be assumed for 

unmonitored workers [14]. For 1994 forward, 100 mrem/yr should be assumed [14]. This dose 

should be treated as photons in the 30- to 250-keV energy group (ORAUT 2006b). 

Table 4-6 summarizes external dose to be assigned to monitored and unmonitored workers at West 

Valley to account for excess background subtraction and onsite ambient exposures. All external dose 

assigned from Table 4-6 should be photons in the 30- to 250-keV energy group (ORAUT 2006b). 

Table 4-6. External onsite environmental dose for West Valley workersa

. 

Year

Monitored workers Unmonitored workers

Additional dose 

(mrem/yr) Uncertainty

Ambient dose 

(mrem/yr) Uncertainty

1965 312 Normal, ±30% N/A N/A

1966 312 Normal, ±30% N/A N/A

1967 312 Normal, ±30% N/A N/A

1968 312 Normal, ±30% N/A N/A

1969 312 Normal, ±30% N/A N/A

1970 312 Normal, ±30% N/A N/A

1971 312 Normal, ±30% N/A N/A

1972 312 Normal, ±30% N/A N/A

1973 312 Normal, ±30% N/A N/A

1974 18 Constant N/A N/A

1975 18 Constant N/A N/A

1976 18 Constant N/A N/A

1977 18 Constant N/A N/A
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 59 of 152

Year

Monitored workers Unmonitored workers

Additional dose 

(mrem/yr) Uncertainty

Ambient dose 

(mrem/yr) Uncertainty

1978 18 Constant N/A N/A

1979 18 Constant N/A N/A

1980 18 Constant N/A N/A

1981 18 Constant N/A N/A

1982 0 N/A 170 Constant

1983 0 N/A 170 Constant

1984 0 N/A 170 Constant

1985 0 N/A 170 Constant

1986 0 N/A 170 Constant

1987 0 N/A 170 Constant

1988 0 N/A 170 Constant

1989 0 N/A 170 Constant

1990 0 N/A 170 Constant

1991 0 N/A 170 Constant

1992 0 N/A 170 Constant

1993 0 N/A 170 Constant

1994 0 N/A 100 Constant

1995 0 N/A 100 Constant

1996 0 N/A 100 Constant

1997 0 N/A 100 Constant

1998 0 N/A 100 Constant

1999 0 N/A 100 Constant

2000 0 N/A 100 Constant

2001 0 N/A 100 Constant

2002 0 N/A 100 Constant

2003 0 N/A 100 Constant

2004 0 N/A 100 Constant

2005 0 N/A 100 Constant

2006 0 N/A 100 Constant

2007 0 N/A 100 Constant a Additional dose should be prorated as appropriate for partial years of employment.

5.0

5.1 INTERNAL DOSE CONTROL PROGRAM

INTERNAL DOSIMETRY

5.1.1

The internal dose control program at West Valley during operations consisted of screening methods 

such as air samples and nasal swabs as a method to estimate internal exposures. Bioassay was 

viewed as an evaluation aid in determining historical levels of exposure, and not as a means for 

exposure control (O’Reilly 1975). The screening tools were based on the legal thresholds for 

mandatory personnel monitoring, i.e., 25% of the applicable AEC or NRC limits (O’Reilly 1975). NFS 

was under regulation and oversight from the AEC over the entire operations era. Regulatory authority 

transitioned to the NRC in 1974. Radiation protection standards for the site were therefore prescribed 

by 10 CFR Part 20 throughout the operations and residual radiation eras, until operational authority 

was transferred to DOE on February 26, 1982. 

Operations and Residual Radiation Eras 1966 – February 25, 1982
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 60 of 152

5.1.1.1 Nasal Swabs

As of 1968, and through at least 1974, NFS’ action limits for nasal swabs were 50 dpm alpha and 

5000 dpm beta (O’Reilly 1975). NFS stated nasal swab results were “used as a screening technique 

to determine when further evaluation is required. If the nasal swabs indicate the alpha activity is less 

than 50 dpm, the exposure is considered by NFS to be far enough below the limits specified in 10 

CFR 20.103 that further evaluation is not required.” They asserted their 50 dpm alpha criterion 

corresponded to “about 25%” of the Part 20 limit, and point out that personnel monitoring was not 

required until 25% of the limit was likely to be exceeded. “This specific procedure has been in effect 

at NFS since 1968” (O’Reilly 1975). 

As of 1972, nasal swabs were taken after “each zone entry” and counted for alpha and beta. If 

activity was detected, urinalysis and a whole-body count were performed. If alpha activity above 

3 dpm/d was detected in the urinalysis, then fecal sampling was performed (Nelson 1973a, p. 5 of 

attachment). 

NFS’ 1973 Safety Analysis Report says nasal swabs were obtained upon exiting from Zone 4 areas to 

screen for inhalations (NFS 1976). 

1974: Nasal swabs were taken when “potential for inhalation of radioactive materials has increased.” 

“Based on the activity found on nasal swabs, additional internal monitoring such as bioassay sampling 

or whole-body counting may be required” (NFS 1974b). 

5.1.1.2 Chest Counter

The chest counter was a 3- by 3-in. NaI detector formerly used as a thyroid monitor. Per Keely and 

Wenstrand (1971), the detector was collimated and used at a distance of 10 cm from the chest. (The 

detector was at 10 cm when the collimator was in contact with the subject being counted.) The chest 

counter was used at West Valley as a screening tool. Initially comparisons were made between the 

gross count rate from an individual suspected of an intake and that from a control subject. Later the 

procedure evolved to use a phantom as a control and a conversion factor for computing a measure of 

intake. The chest counter was in use at least as early as November 1966 (Helgeson 1966), and was 

used routinely back to at least 1967 (Miller 1972a). 

As of 1971, the NFS procedure for use of the chest counter was to align the system using 137Cs, count 

a phantom, then count the individual of interest. The counts were performed with the collimator in 

contact with the subject. The gross count rates for a region of interest 15 channels (8 below and 7 

above) about the 137Cs peak were determined for the phantom and the individual. The net count rate 

was computed and multiplied by a conversion factor to get “relative lung burden” (HNS 1971). 

As of September 1972 the chest counter had been replaced by an on-site whole-body counter (Miller 

1972a). 

5.1.2

Presently, the WVDP uses a combination of bioassay and air sampling data for internal dose 

assessment. Air sampling data are used in cases where bioassay methods lack requisite sensitivity, 

e.g., for the alpha constituents of the mixed fission and activation product source term. In these 

cases, the detection limits associated with bioassay methods are such that the corresponding dose 

could exceed the site investigation level if it is assumed the date of intake was 12 months earlier. 

Derived air concentration (DAC)-hour calculations are therefore used to assess exposures to airborne 

DOE Era February 26, 1982 – present
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 61 of 152

alpha activity. Workers for whom decision levels are established via alpha air monitoring must also 

participate in the alpha in vitro bioassay program (Kubiak 1998 p. 14). 

The WVDP investigation level for internal dose assessment is 100 mrem committed effective dose 

equivalent (CEDE). If a DAC-hour calculation gives a CEDE of 10 mrem or less, no additional action 

is taken. However, all data and calculations are placed in the worker’s dosimetry file. If the calculated 

CEDE exceeds 10 mrem, then additional sampling is required (Kubiak 1998 p. 40). 

5.2 BIOASSAY PROGRAM

5.2.1

In vitro bioassay at West Valley during its operations era was a combination of routine and for cause 

analyses. Records show that ad hoc urine and fecal analyses were performed at least as early as 

September 1967. A routine urinalysis program was in place since at least 1968. It appears a routine 

fecal analysis program started sometime between 1968 and 1970 (Wenstrand 1968c) [15]. 

Operations and Residual Radiation Eras 1966 – February 25, 1982

In vivo bioassay at West Valley during its operations era was performed using the chest counter 

(described above), a commercial mobile whole-body counting service, and a phoswich counter 

operated by the New York University (NYU) Medical Center in Rochester, New York. The latter was 

used for follow-up in vivo counts for TRU elements following significant intakes. An onsite in vivo

counter was installed at West Valley sometime in 1972 (Miller 1972a). 

As of 1967, the NFS bioassay program for plutonium depended upon the location in the plant where 

an intake may have occurred. If a suspected intake occurred in a mechanical processing area, it was 

assumed the intake would have been accompanied by fission products. Therefore, for these areas, 

the internal dose control program consisted of a combination of nasal swabs to look for gross beta 

activity and in vivo counting using the NFS chest counter. Follow-up bioassay (fecal sampling or in 

vivo counting) would be ordered based on the results of the nasal swab and chest count. Plutonium 

in the mechanical processing (head end) areas would most likely have been in an insoluble form 

(metal or oxide) (Runion 1967a, p. 3).

If a possible plutonium intake occurred in a chemical processing area, the plutonium would have been 

in nitrate (soluble) form. Urinalysis was therefore used for these individuals (Runion 1967a). 

In April 1968 Clark (1968b) asserted NFS’ procedures required urine samples if: 

• There was a positive nasal swab

• Work was performed in an area with alpha contamination greater than 500,000 dpm/100 cm2

• There was a loss of air supply in an area with smearable alpha contamination greater than 500 

dpm/100 cm2 or smearable beta contamination greater than 50,000 dpm/100 cm2

• Facial contamination was detected above 50 cpm alpha or 100 cpm beta (per probe area)

It is assumed that a “positive” nasal swab referred to a result in excess of the action limits given in 

Section 5.1.1.1. 

As of 1968 fecal sampling was not part of the NFS bioassay program except for on an ad hoc basis 

when deemed necessary by their Health and Safety Department (Clark 1968b). However, a February 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 62 of 152

29, 1968, memorandum discusses start-up of a fecal monitoring program for workers whom routinely 

entered alpha-contaminated areas (Wenstrand 1968c). By 1970 this routine fecal sampling was an 

element of their program. A summary report from the bioassay technician for 1970 says 200 fecal 

samples were shipped out that year (Kester 1971). 

As of 1972, “All permanent employees exposed to radiation receive an annual physical from the 

plant’s physician” which included blood counts and “checks for internal alpha (plutonium) exposure” 

(Johnson & Higgins 1972]. NFS (1976) says a baseline radiation history was established for each 

radiation worker, which included a physical, blood counts, and urinalysis records. Blood counts and 

urinalysis were repeated semi-annually. It is not known how far back this practice dated. 

As of September 1972 NFS had replaced the chest counter with a “shielded whole-body counter”, and 

stated that “all plant personnel are now routinely counted by the whole-body counter once yearly and 

at any other time there is reason to believe an uptake has occurred” (Nelson 1973a, p. 5 of 

attachment). The annual whole-body counts replaced urinalyses for mixed fission products under an 

assumption that intakes of 90Sr or 147Pm would be accompanied by other, gamma-emitting isotopes 

(Nelson 1973b, p. 8 of enclosure).

NFS (1976) states urine samples were collected semi-annually on all radiation workers “to evaluate 

the potential long-term buildup of plutonium in the body.” It appears this practice was in place at least 

back to 1969 (e.g., see Keely and Wenstrand 1971). Fecal samples were also collected from 

radiation workers on an annual basis and analyzed for plutonium. It appears this practice began 

sometime between 1968 and 1969 (see Clark 1968b and Keely and Wenstrand 1971).

As of October 1973 the frequency for plutonium urinalysis had been decreased to annually for 

radiation workers (Nelson 1973b, p. 8 of enclosure). NFS’ calculations showed the program would 

detect a single intake of soluble plutonium of 20% of a MPBB 12 months post-intake. “Uranium 

exposures are evaluated by using plutonium as a tracer.” “Commercial waste burial contractor 

employees” received annual whole-body counts and annual urinalyses for mixed fission products and 

tritium (Nelson 1973b, p. 9 of enclosure). Tritium had accumulated in water in burial trenches up to 

0.1 μCi/cm3

. In an inspection report the AEC remarked on the fact annual urinalysis for tritium would 

represent a high limit of detection. Other contractor employees received bioassay evaluations if 

“nasal smears are positive and exceed specified limits” (Nelson 1973b, p. 9 of enclosure). 

As of 1974 the internal monitoring program for radiation workers in the process building remained as it 

was in about October 1973. Annual urine samples were collected and analyzed for plutonium and all 

radiation workers were whole-body counted annually for fission products (NFS 1974b). 

A May 18, 1977, memorandum says NFS’ procedures had been modified to implement entrance and 

termination whole-body counts. This was apparently not being done before this time (Oldham 1977). 

NFS sometimes experienced delays in processing bioassay samples. A December 28, 1967, letter 

from the AEC stated, “The bioassay program needs modification and improvements. Specifically the 

procedure for issuance, follow-up and retrieval of filled urinalysis sample bottles from employees as 

well as records thereof have not been effective in some instances ... A better effort is necessary in 

obtaining more timely samples following evidence of possible internal depositions.” Earlier that year 

(September), 38 bioassay samples were lost due to an error by the parcel delivery service used (Clark 

1968b). In 1974 a worker involved with possible inhalation exposure submitted urine samples on 

February 23 and 24 which were not sent to the vendor for analysis until May 16, 1974 (O’Reilly 1975). 

It is not known how NFS stored its samples or if preservatives were added. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 63 of 152

5.2.1.1 Bioassay Records

As of April 1968 a “Bioassay Log” was kept documenting the names of employees involved in the 

bioassay program and dates their samples were issued and shipped (Clark 1968b). 

As of April 1972 exposure to airborne radioactive material was recorded on a “Lapse of Radiation 

Control” form. “Bioassay results and whole body counts are recorded on appropriate forms in the 

individual personnel folders and are available for examination by the employee” (Miller 1972a). A 

September 7, 1972, memorandum confirms this practice (Duckworth 1972). 

In its 1973 Safety Analysis Report, NFS said dosimetry records were maintained by the Health and 

Safety Department for each employee or worker (NFS 1976). These records included previous 

radiation exposure, current radiation exposure, urinalysis results, fecal analysis results, whole-body

counting data, and “special reports,” such as exposure estimates for lost or damaged film badges. 

These reports were submitted to the individuals and to the AEC as required. 

5.2.2

Bioassay at the WVDP comes under six categories (Kubiak 1997): 

DOE Era February 26, 1982 - present

• Baseline upon initial employment

• Re-entry, upon resumption of bioassay for an employee previously monitored

• Periodic routine bioassay

• Termination bioassay upon termination of employment or radiation worker status

• Discontinue bioassay upon transfer from monthly to quarterly external monitoring or removal 

from the respiratory protection program (in vitro only)

• Special bioassay

5.2.2.1 Routine Bioassay

The routine bioassay program at the WVDP includes both in vivo and in vitro monitoring. 

All radiation workers are included in the in vivo bioassay program. In addition, any female whom 

declares a pregnancy receives both in vivo and in vitro bioassay at the time of declaration and when 

the pregnancy is declared ended (Kubiak 1998, p. 35).

All dosimetry qualified personnel monitored on a monthly basis or individuals with respiratory 

protection qualifications for work in radiological areas are included in the WVDP in vitro bioassay 

program. The in vitro program required baseline, annual, and termination sampling (Kubiak 1998, 

p. 31). 

As of July 31, 1990, routine in vitro analyses for the WVDP were to include (WVNSC 1990a): 

• Plutonium isotopic

• Total uranium

• 241Am
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 64 of 152

• 90Sr

As of December 7, 1990, routine in vitro analyses for the WVDP were to include (WVNSC 1990b): 

• Plutonium isotopic

• Total uranium

• 90Sr

5.2.2.2 Special Bioassay

Special bioassay is required if action levels established for workplace surface and airborne 

contamination levels and personnel contamination are reached or exceeded, or if otherwise requested 

by the RP Department management or technical staff. Special bioassay can be in vivo and/or in vitro. 

The action levels for special bioassay are (Kubiak 1998, p. 33): 

• A routine bioassay measurement is determined to be positive for radioactive content.

• Facial contamination or detectable radioactivity on nasal or mouth swabs indicates a potential 

for intake. 

• A respiratory protection factor for a respirator is exceeded and exposes the worker to airborne 

concentrations exceeding the DAC.

• Airborne monitoring indicates the potential for internal dose in excess of 100 mrem CEDE. 

• Wound contamination (in or around a skin break), which is: 

– Greater than 100 dpm/100 cm2 alpha

– Greater than 10,000 dpm/100 cm2 beta/gamma

– Extensive personnel decontamination is required

5.2.3

Table 5-1 gives a summary of ad hoc and routine in vivo monitoring carried out at West Valley from 

1966 to the present.

Summary

Table 5-1. Summary of in vivo bioassay frequencies for West Valley radiation workers. 

Period Ad hoc Routine

1966–September, 1972 • Chest counter used for onsite screening

• Ad hoc counts for TRU using the 

phoswich counter at NYU-Rochester 

(post-incident follow-ups)

• Commercial, mobile whole body counter for 

annual counts of selected workers

• Routine counts using the chest counter 

(frequency unclear)

September, 1972–1977 • Onsite whole body counter used for 

screening of suspected intakes 

involving MFP

• Radiation workers were counted annually for 

MFP using onsite whole body counter

1977–presenta • Onsite whole body counter used for 

screening of suspected intakes 

involving MFP

• Radiation workers were given baseline, 

annual, and termination counts for MFP 

using onsite whole body counter

a. In 1993 the system changed from a NaI detector and a bed geometry to an intrinsic germanium detector in a stand-up geometry.

Table 5-2 gives a summary of routine in vitro monitoring carried out at West Valley from 1966 to the 

present. In vitro bioassay was also performed ad hoc based on nasal smears or other indicators of 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 65 of 152

possible intake. The type of bioassay (urinalysis or fecal analysis) was selected based on the 

chemical form (soluble or insoluble) of suspected material. 

Table 5-2. Summary of routine in vitro bioassay frequencies for West Valley radiation workers. 

Period Routine urinalysis Routine fecal analysis

1966–1972 • Semiannual for all radiation workers for plutonium

• ≥ Annual for selected workers for MFP (may have 

been more frequent than annually)

Annual for all radiation workers 

beginning ca. 1968.

1973–February, 1982 • Annual for all radiation workers for plutonium

• Annual for MFP and tritium for SDA workers 

(through 1975)a, b

None known

March, 1982–late 1990 • Baseline, annual, and termination for workers with 

monthly TLD exchange or with respiratory protection 

qualifications. Analytes were plutonium isotopic, 

total uranium, Am-241 and Sr-90. 

None known

Late 1990–present • Baseline, annual, and termination for workers with 

monthly TLD exchange or with respiratory protection 

qualifications. Analytes were plutonium isotopic, 

total uranium, and Sr-90. 

None known

a. Annual whole body counts replaced routine urinalysis for mixed fission products (MFPs) for workers in the process building.

b. The SDA was closed in May 1975.

5.3 IN VITRO ANALYSES

5.3.1

In vitro bioassay at West Valley during its operating era was generally performed by commercial 

vendors. However, it appears nuclide-specific urinalyses (i.e., those not analyzed for mixed fission 

products) were performed in house in 1970 (Kester 1971). It is not known when this practice actually 

began or how long it continued. 

Operations and Residual Radiation Eras 1966 – February 25, 1982

As of June 1967 the minimum detectable “concentration” [sic] for the plutonium urinalysis vendor was 

0.45 dpm/d (Clark 1968c). 

A letter dated March 19, 1968, states the minimum detectable excretion rate for the bioassay

laboratory performing their urinalyses for plutonium was “0.3 dpm/L”

NFS did not normalize its urinalysis results to a reference excretion rate of 1.4 L/d. Activity per 24 

hours was computed by taking the product of the analysis result in activity per unit volume and the 

total volume collected and dividing the resulting activity by the sampling interval (e.g., see Author 

unknown 1974). So for example if 0.85 L was collected over a period of 12 hours and showed 

0.12 dpm/L, the result would be recorded as 

[sic] (Monroe 1968). Keely and 

Wenstrand (1971) show a value of 0.3 dpm/d for plutonium nitrate in urine. 

dpm/d 0.20 d 0.5

L 0.85 dpm/L 12.0 = ×

The same result collected over 24 hours would have been reported as 0.10 dpm/d. 

In general 24-hour urine samples collected at West Valley during the operations and residual radiation 

periods had an excretion volume of less than 1 L. Some instances were noted of greater than 1 L 

collections for periods greater than 24 hours in the NOCTS files, but these were uncommon. Users 

should therefore multiply urinalysis results reported in terms of activity excreted per liter by 1.4 to 

correct them to a reference excretion rate of 1.4 L/d. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 66 of 152

The “Yearly Report from Bioassay Technician” for 1970 implies urinalyses for other than mixed fission 

products were performed onsite, and that counts were often lost “due to electrical storms.” His 

numbers show approximately 10% of the counts were lost. The onsite counting in 1970 included 

samples collected in 1969 since they lost three months of counting due to a strike. Counting began in 

April 1970. For the year they counted 393 urine samples in house, shipped out approximately 150 

samples for urinalysis for mixed fission products, and shipped out approximately 200 fecal samples 

(Kester 1971). 

Sometimes a batch of urine samples from an individual would be submitted to the analysis lab with 

instructions to analyze the most recent sample first. The remaining samples were only to be analyzed 

if the initial analysis showed an activity concentration in excess of some threshold. Otherwise the lab

was instructed to discard the remaining samples. 

As of 1973 plutonium urinalysis was performed by a commercial vendor via plutonium separation and 

alpha spectroscopy. It was not performed onsite. The detection limit was 0.2 dpm/L (Nelson 1973b, 

p. 9 of enclosure). NFS’ calculations showed the program would detect a single intake of soluble 

plutonium of 20% of a MPBB 12 months post-intake. “Uranium exposures are evaluated by using 

plutonium as a tracer.”

In its report summarizing an inspection performed in October 1973 the AEC assumed a detection limit 

for tritium urinalysis of 0.02 μCi/L. It is not explicitly clear this was a representative MDC (Nelson 

1973b, p. 9 of enclosure).

Table 5-3 summarizes analytes, vendors, and observed reporting levels for in vitro analyses at West 

Valley during the operations and residual radiation eras. The reporting levels represent “less than” 

values given in analysis reports, e.g., “< 0.3 dpm/L.” In many cases reporting levels may be inferred 

from the analysis reports from the various vendors. Some reporting levels are a priori values, others 

are a posteriori. Reporting levels are not apparent in analysis reports for samples associated with 

incidents or which otherwise showed positive activity for all samples. 

The information in Table 5-3 comes from NOCTS and other records. The analyses listed should not 

be interpreted as complete or inclusive. Analytes in addition to those in the table are likely. 

Table 5-3. In vitro analyses and reporting levels for West Valley during the 

operations and residual radiation eras.

Year Analyte Vendor Observed reporting levels

1967

MFP in urine Controls for Radiation, Inc. 10 to 50 dpm/L

Gross beta in urinea Teledyne 7770 dpm/Lb

Total U in urine Controls for Radiation, Inc. 0.1 μg U/L

Teledyne 0.1 μg U/L

Total Pu in urine Controls for Radiation, Inc. 0.2 to 0.4 dpm/L

Teledyne 0.3 dpm/L

1968

MFP in urine Teledyne 30 dpm/L

Total U in urine Teledyne 0.1 μg U/L

Total Pu in urine Teledyne 0.3 dpm/L

Pu-239 in feces Eberline 0.9 dpm/samplec

Total Pu in feces Teledyne 0.3 dpm/g

Eberline 0.3 dpm/gd

1969

MFP in urine unknown 50 dpm/Le

Gross beta in urine Eberline 7770 dpm/Lb

Sr-90 in urine Eberline 444 dpm/Lb

U-232 in urine unknown 0.22 dpm/Lf
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 67 of 152

Year Analyte Vendor Observed reporting levels

U-235 in urine Eberline 0.22 dpm/Lf

Total Pu in urine unknown 0.3 dpm/Lg

MFP in feces unknown 22 dpm/sampleh

Pu-239 in feces Eberline 0.9 dpm/samplec

Total Pu in feces Eberline 0.3 dpm/gd

1970

MFP in urine unknown 50 dpm/Le

Pu-239 in feces Eberline 0.9 dpm/samplec

Total Pu in feces Eberline 0.3 dpm/gd

1971 MFP in urine unknown 50 dpm/Le

Total Pu in feces unknown 0.3 dpm/gd

1972

MFP in urine unknown 50 dpm/Le

Pu-239/240 in urine Ledoux & Company 0.09 to 0.20 dpm/L

Total Pu in urine Ledoux & Company 0.20 to 0.40 dpm/L

Total Pu in feces unknown 0.3 dpm/gd

1973 Pu-239/240 in urine unknown 0.3 dpm/Lg, i

Total Pu in urine unknown 0.3 dpm/Lg, i

1974

MFP in urine Ledoux & Company 50 dpm/Le

Tritium in urine Ledoux & Company 4E-3 μCi/L

Pu-238 in urine Ledoux & Company 0.07 to 0.58 dpm/L

Pu-239/240 in urine Ledoux & Company 0.03 to 0.43 dpm/L

Total Pu in urine Ledoux & Company 0.07 to 0.72 dpm/L

1975

Tritium in urine Ledoux & Company 4E-3 μCi/Lj

Pu-238 in urine Ledoux & Company 0.05 to 0.10 dpm/L

Pu-239/240 in urine Ledoux & Company 0.04 to 0.07 dpm/L

Total Pu in urine Ledoux & Company 0.09 to 0.16 dpm/L

1976

MFP in urine Eberline 50 dpm/Le

Tritium in urine Ledoux & Company 2E-3 μCi/L

Pu-238 in urine Ledoux & Company 0.06 to 0.10 dpm/L

Pu-239/240 in urine Ledoux & Company 0.05 to 0.08 dpm/L

Total Pu in urine Ledoux & Company 0.07 to 0.14 dpm/L

Pu-238 in urine Eberline 0.04 to 0.10 dpm/sample

Pu-239 in urine Eberline 0.05 to 0.07 dpm/sample

1977

Pu-238 in urine Ledoux & Company 0.07 to 0.58 dpm/Lk

Pu-239/240 in urine Ledoux & Company 0.03 to 0.43 dpm/Lk

Total Pu in urine Ledoux & Company 0.07 to 0.72 dpm/Lk

a. Result includes a value of 2 to 3 dpm/L for average 40K excretion for an unexposed individual. 

This implies 40K was not separated prior to counting for gross beta.

b. Selected from ORAUT-TKBS-0026-5 Rev. 0 (technical basis document (TBD) for occupational 

internal dose for Argonne National Laboratory – West) after a review of TBDs showed it to be the 

highest value for the time period of interest. This is a minimum detectable concentration (MDC) 

rather than a reporting level.

c. Selected from ORAUT-TKBS-0007-5 Rev. 1 PC-1 (TBD for occupational internal dose for Idaho 

National Engineering Laboratory) after a review of TBDs showed it to be the highest value for the 

time period of interest. This is an MDA rather than a reporting level.

d. Assumed same as reported by Teledyne. 

e. Assumed high end of range reported by Controls for Radiation, Inc. in 1967. This bounds the 30 

dpm/L value reported by Teledyne in 1968. 

f. Reflects the MDCs given in ORAUT-TKBS-0026-5 Rev. 0 (TBD for occupational internal dose for 

Argonne National Laboratory – West) and ORAUT-TKBS-0007-5 Rev. 1 PC-1 (TBD for 

occupational internal dose for Idaho National Engineering Laboratory) for uranium isotopic 

analyses. 

g. Assumed the value given by Teledyne in 1968, which should be bounding for later analyses. 

h. Reflects the MDA given in ORAUT-TKBS-0007-5 Rev. 1 PC-1 (TBD for occupational internal 

dose for Idaho National Engineering Laboratory) for 60Co, 90Sr and 134Cs. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 68 of 152

i. (Nelson 1973b, p. 9 of enclosure) reports a minimum detectable level of 0.2 dpm/L for plutonium 

in urine in 1973. 

j. Assumed value reported for 1974.

k. Assumed range reported for 1974.

5.3.2

The WVDP continues to use vendors for analysis of in vitro bioassay samples. Required detection 

levels are specified by the WVDP for different analytes based on the requested turnaround time. 

DOE Era February 26, 1982 - present

Table 5-4 shows the WVDP required detection levels for in vitro analyses as of September 12, 1990, 

for different turnaround times. “Priority 1 Processing” means immediate turnaround, Priority 2 was 

approximately 30 days, and routine processing was approximately 3 months (WVNSC 1990c). 

Table 5-5 shows the WVDP required detection levels for priority and routine in vitro samples that were 

to be implemented “on or about” August 1, 1997 (Kubiak 1997, pp. 42–43 ).

Table 5-4. Required detection levels for WVDP in vitro analyses as of September 12, 1990.

Analytical 

method Analytes

Priority 1 processing Priority 2 processing

Routine 

processing

Urine 

(dpm/sample)

Fecal 

(dpm/sample)

Urine 

(dpm/sample)

Fecal 

(dpm/sample)

Urine 

(dpm/sample)

Pu isotopic Pu-238/239/240 8.0E-02 6.0E+01 2.0E-02 6.0E+01 2.0E-02

Am-241 Am-241 8.0E-02 3.3E+01 2.0E-02 3.3E+01 2.0E-02

Cm isotopic Cm-243/244/245 1.2E+00 3.4E+01 2.0E-02 3.4E+01 2.0E-02

Th isotopic Th-228/230/232 1.0E-01 3.0E+00 1.0E-01 3.0E+00 NA

Sr-90 Sr-90 3.4E+01 2.2E+04 3.4E+01 2.5E+04 3.4E+01

Sm-151 Sm-151 1.2E+02 3.1E+05 1.2E+02 3.1E+05 NA

Eu isotopic Eu-154/155 8.0E+00 2.2E+04 8.0E+00 2.2E+04 NA

Total uranium Total U 1.0 μg/sample 240 μg/sample 1.0 μg/sample 240 μg/sample 1.0 μg/sample

U isotopic U-233/234/235/238 1.0E-01 3.0E+01 9.0E-02 3.0E+01 NA

Table 5-5. Required detection levels for priority and routine in vitro samples as of 

August 1, 1997. 

Analytical 

method

Urinalysis Fecal analysis

Priority Routine Priority Routine

Pu isotopic 2.0E-01 dpm/L 2.0E-02 dpm/L 6.0E+01 dpm/sample N/A

Am-241 2.0E-01 dpm/L 2.0E-02 dpm/L 3.3E+01 dpm/sample N/A

Th isotopic 1.0E-0 dpm/L N/A 3.1E+00 dpm/sample N/A

Sr-90 8.9E+00 dpm/L 1.1E+00 dpm/L 3.1E+00 dpm/sample N/A

Total U 2.0E-04 μg 2.0E-04 μg 5.3E+08 dpm/sample N/A

U isotopic 1.0E-01 dpm/L N/A 3.1E+01 dpm/sample N/A

Table 5-6 shows the WVDP required detection levels for in vitro analyses as of March 14, 1998, for 

different turnaround times (WVNSC 1997). 

Table 5-6. Required detection levels for WVDP in vitro analyses as of March 14, 1998.

Analytical 

method Analytes

Urinalysis (dpm/L) Fecal analysis (dpm/g)

Priority Routine Follow-up Priority Routine

Pu isotopic Pu-238 2.2E-02 2.2E-02 2.1E-02 2.2E-03 N/A

Pu-239/240

Sr-90 Sr-90 8.9E+00 1.1E+00 9.1E-01 2.2E-02 N/A

KPAa Total uranium 4.4E+05 4.4E+05 4.4E+05 N/A N/A
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 69 of 152

Am-241 Am-241 2.2E-02 2.2E-02 2.1E-02 2.2E-03 N/A

Th isotopic Th-228 2.2E-02 2.2E-02 2.1E-02 3.8E-03 N/A

Th-230

Th-232

U isotopic U-233/234 2.2E+01 3.3E-02 3.3E-02 4.4E-03 N/A

U-235

U-238

Gamma spec Cs-137 3.8E+01 3.8E+01 N/A N/A N/A

a. KPA = kinetic phosphorescence analysis. 

As of September 12, 1990, if WVDP received an in vitro result showing a concentration greater than 

the required detection level, an evaluation was performed to determine if the result should be 

considered positive. A result greater than three times the 1-sigma error was considered positive and 

a confirmatory sample was collected. If the confirmatory sample also showed positive activity, then a 

dose assessment was performed. Results less than three times the 1-sigma error were considered 

nondetections (WVNSC 1990c). This practice continued until July 1997. 

In July 1997 WVDP revised its decision criterion for positive in vitro results to one based on the 

standard deviation observed for a series of blank samples. An in vitro result greater than two times 

the standard deviation for the representative blank was considered positive (Kubiak 1997). This was 

subsequently revised further to assert decision levels as a fraction of minimum detectable activity 

(MDA). The WVDP decision levels as a fraction of MDA are given in Table 5-7. These values are as 

of November 23, 1998 (Kubiak 1998). 

Table 5-7. WVDP decision levels for positive in vitro samples as of 

November 23, 1998. 

Analysis Urine detection limit Fecal detection limit

Sr-90 ≥ 0.93 MDA and ≥ 4.9E-10 μCi/mL Not performed

Total uranium ≥ 3.8E-05 μg/mL Not performed

Pu isotopic Pu-238 ≥ 0.77 MDA Pu-238 ≥ 0.71 MDA

Am-241 ≥ 0.71 MDA Not performed

Th isotopic Any Th result ≥ 0.71 MDA Not performed

U isotopic U-234 and U-238 ≥ 0.71 MDA Not performed

U-234/U-238 > 3 or

U-234 ≥ 0.71 MDA

U-238 < 0.71 MDA

5.4 IN VIVO ANALYSES

5.4.1

5.4.1.1 Chest Counter

Operations and Residual Radiation Eras

In a December 1966 letter Helgeson states he felt that the NFS chest counter would see levels “much 

less” than 1 μCi for typical fission and activation products, and most likely levels “less than a few tens 

of nanocuries” (Helgeson 1966). 

Results from the NFS chest counter are sometimes reported in nanocuries of 137Cs and fraction of 

maximum permissible lung burden (MPLB). Other reports of chest counter data show total counts, a 

background count, the net count rate, and "relative lung burden" computed from the net count rate. 

Example values of relative lung burden are 20.17 and 13.17. Relative lung burden may be the same 

as percent MPLB, but this has not been verified. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 70 of 152

5.4.1.2 Commercial Mobile Whole-Body Counting

December 1, 1966

Two West Valley workers were counted on December 1, 1966. Positive activity was reported for 40K, 

60Co, 95Zr/Nb, and 137Cs for both individuals. Positive 106Ru/Rh was reported for one of them. No 131I 

was observed in either individual (Helgeson 1966). 

April 17 to 20, 1967

Helgeson Nuclear Services performed whole-body counts on 160 radiation workers at West Valley 

April 17 to 20, 1967 (Wenstrand 1967c). “Trace quantities of gamma emitting nuclides other than 

potassium-40 and cesium-137 were detected in 22% of the individuals counted. The maximum 

deposition was from Zr/Nb-95.” Nuclides identified were 60Co, 65Zn, 58Co (134Cs), 95Zr/Nb, 137Cs, and 

103/106Ru. It was stated “All Co-58 results may be interpreted as Cs-134” (HNS 1967). Approximately 

one-third of Production, Maintenance, and Health and Safety personnel showed detectable internal 

deposition of radionuclides (Keely 1967b). 

The mobile whole-body counter used was a scanning bed counter in a 24-ft van with an 8- by 4-in.

NaI detector. Individuals were counted for approximately 8 minutes (HNS 1967). 

Table 5-8 shows sensitivity data for the April 1967 whole-body counting for “typical” background at 

2.326 sigma (99% confidence) (HNS 1967).

Table 5-8. Sensitivity data for April 1967 in vivo

counting. 

Isotope Background (cpm) MDA

K-40 72.4 8.8 g

Co-60 38.7 0.81 nCi

Zn-65 57.8 2.73 nCi

Co-58 73.9 1.70 nCi

Zr/Nb-95 93.1 1.28 nCi

Cs-137 129.8 1.89 nCi

Ru-103/106 333.6 2.22 nCi

The format of the reports for the April 1967 in vivo counting is as follows. 

Last name, First name + Middle initial, Social Security Number (SSN) 

Weight in pounds followed by seven values of net count rate (cpm) for 40K, 60Co, 65Zn, 58Co, 

95Zr/Nb, 137Cs, and 103/106Ru, respectively. 

40K result in grams of potassium

60Co result in nanocuries

65Zn result in nanocuries

58Co result in nanocuries

95Zr result in nanocuries

137Cs result in nanocuries

103/106Ru result in nanocuries

Note that reported results may be negative. 

128 people were counted by the mobile whole-body counting vendor from April 15 to 17, 1968. “The 

very high background encountered caused severe disturbances in the low energy end of the 

spectrum, below approximately 0.4 MeV.” This affected the sensitivity for nuclides with primary 

April 15 to 17, 1968
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 71 of 152

gamma lines below this energy, such as 131I, 51Cr, and 141/144Ce. The vendor estimated, based on 

visual inspection of the spectra, that none of those isotopes was present at a level greater than or 

equal to 10% of the MPBB. “The background at the whole-body counter was the highest in our 

experience.” This was said to render their simultaneous equations for evaluating results “useless” 

(HNS 1968). 

The format for 1968 in vivo counting reports is as follows (HNS 1968).

Name and SSN

Results: nuclide, units; result; two sigma confidence level; percent of MPBB or lung burden if 

determined to be insoluble. For potassium, the last value is the ratio of the grams of 

potassium to the individual’s body weight in kilograms.

The commercial provider was on site for mobile whole-body counting from April 17 to 18, 1969. Only 

29 people were counted. “Low levels of typical mixed fission and corrosion products were seen in a 

number of people” (HNS 1969a). “Eight people were measured for americium-241 and small 

photopeaks were visualized in the graphs in each of the eight. In every case, however, the level of 

activity was below what we define as our minimum sensitivity of 0.5 nanocurie.” One individual was 

also counted for 235U. The 241Am and 235U counts were performed using a 0.5- by 8-in. NaI detector. 

These were lung counts as opposed to whole-body counts (HNS 1969a). The sensitivity for 235U was 

asserted to be 50 μg (HNS 1969a). The 50-μg sensitivity for 235U was asserted to be at a confidence 

level of “about 80%” (HNS 1969b). The 241Am sensitivity is said to be 0.5 nCi ±0.5 nCi at 2 sigma. 

April 17 to 18, 1969

The in vivo counts (whole body and lung) for April 1969 were performed at the “schoolhouse,” which 

afforded a lower background than that experienced in previous years (HNS 1969a). 

The results reports for 1969 should be self-explanatory. Table 5-9 shows sensitivity data for an 

8-minute count using a background measured April 17, 1969, and assuming 140 g of potassium plus 

10 nCi of 137Cs in an individual. Sensitivity was defined as three times standard deviation of 

background divided by the calibration factor (HNS 1969a). 

Table 5-9. Sensitivity data for April 1969 in vivo counts. 

Isotope Background (cpm) Sensitivity (nCi)*

Potassium 90.81 22.5 g

Sb-124 41.93 2.3

Sb-125 224.88 7.4

Ba/La-140 56.71 2.0

Ce-144 497.99 27.3

Cs-134 153.32 2.3

Cs-137 186.77 2.6

Cr-51 351.97 29.0

Co-58 147.61 2.3

Co-60 82.12 1.7

Au-198 335.80 3.0

I-131 282.42 2.4

Ir-192 253.41 2.1

Mn-54 150.29 2.1

Hg-203 547.64 5.1

Ru-106 299.41 12.7

Se-75 247.67 5.2
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 72 of 152

Ag-110m 126.29 2.2

Sr-85 299.41 2.9

Ta-182 118.08 2.4

Tc-99m 473.99 1.9

Sn-113 185.98 3.0

Zn-65 103.94 4.5

Zr/Nb-95 167.63 2.7

May 10 to 11, 1971

The commercial provider was on site for mobile whole-body counting from May 10 to 11, 1971. Only 

11 people were counted (HNS 1971). It appears they were all counted mixed fission and activation 

products (MFAPs), 241Am, and 239Pu. Two of the 11 persons showed positive lung burdens of 

239Pu/241Am. The counts for MFAPs were still performed using the scanning bed counter. The 

239Pu/241Am counts were lung counts performed using a pair of 5-in.-diameter dual-scintillator

(phoswich) detectors. 

The “sensitivities” reported for mixed fission and activation products for the May 1971 in vivo counting 

were essentially the same values as from 1969 (HNS 1971). Sensitivities for plutonium counts were 

asserted for each individual based on estimates of chest wall thickness.

July 17 to 18, 1972

The commercial provider was on site for mobile whole-body counting July 17 to 18, 1972. Eleven 

individuals were lung counted for 239Pu and 241Am, and four were counted for MFAPs (whole body). 

Due to a miscommunication, the persons counted for TRU elements were not counted for MFAPs, 

though they were supposed to have been. The lung counts for TRU elements were performed using 

the same phoswich detectors used in 1971. A 0.5- by 8-in. NaI was also used “for some studies in the 

lower energy regions” (HNS 1972). The initial results indicated one positive 241Am count and two 

possible plutonium depositions (Wenstrand 1971d). 

The sensitivity values for the July 1972 counts do not differ significantly from those given above for 

the 1969 counting.

5.4.1.3 New York University Medical Center Phoswich Counter

West Valley sometimes sent individuals involved in significant plutonium intakes for follow-up in vivo 

counting using the phoswich system operated by the NYU Medical Center in Rochester, New York. 

The Medical Center had two counters: a 8- by 4-in. NaI and a CsI-NaI phoswich detector. Individuals 

were counted for plutonium only unless sufficient time had elapsed since intake for ingrowth of 241Am. 

Counts were performed of the head, chest, or liver.

In November 1967 the minimum sensitivity of the NYU plutonium counter is said to be an equivalent 

of 8 nCi of plutonium in the whole body based on a skull count (Lewis 1967a). As of July 1968 NYU 

stated the detection limit of their counter for plutonium was 1.0 nCi in the skull, corresponding to 

approximately 10 nCi in the whole body, and 2.0 nCi in the chest or liver (Laurer 1968). 

5.4.2

When DOE took over responsibility for operations at West Valley in February 1982, a bed-type lung 

counter was in use for in vivo monitoring. The counter used a single 4- by 4-in. NaI detector and a 

256-channel MCA. The reported MDAs were 35 nCi for 137Cs and 25 nCi for 60Co. The 4-by-4 

detector was used through at least February 1987. Later versions of in vivo monitoring procedures 

describe the lung counter as a 5- by 3-in. NaI detector surrounded by a 1-in. lead shield. This 

U.S. Department of Energy Era
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 73 of 152

detector was in use by November 1988 (WVNSC 1988). Johnson (1991) reports MDAs for this 

system (for lung counts at contact) of 190 Bq (5.1 nCi) for 137Cs and 55 Bq (1.5 nCi) for 60Co. The 

lung counting system was only calibrated for 137Cs and 60Co. 

As of May 12, 1993, the in vivo monitoring program had transitioned to a high resolution, stand-up 

scanning whole-body counter utilizing hyper-pure germanium detectors (WVNSC 1993). The MDA for 

this system is 7 nCi of 137Cs, which is used as the tracer analyte in the WVDP in vivo monitoring 

program (Kubiak 1997, p. 20). Currently count-specific MDAs are reported for the nuclides in the 

counting library used (e.g., mixed fission and activation products). Hoffman (1997) states the in vivo

counter can detect internal deposition of gamma-emitting nuclides “at a level much less than 10 

nanocuries” and less than 1% ALI. At the time the in vivo laboratory participated in intercomparison 

studies. 

5.5 INTERPRETATION OF BIOASSAY MEASUREMENTS FOR WEST VALLEY WORKERS

Tables 5-10, 5-11, and 5-12 show nuclide mix information for the MFAP, uranium, and plutonium 

source terms at West Valley as of January 1, 2000. These data were derived from (WVNSC 1997) by 

converting CEDE fractions to activity fractions. The alpha activity fractions for the plutonium mixture 

given in Table 5-12 were approximated by removing the 241Pu activity and renormalizing the remaining 

constituents. 

Table 5-10. Nuclide mix information for 

West Valley MFAP source term as of 

January 1, 2000. 

Nuclide Activity fraction

Sr-90 46.73%

Cs-137 51.87%

Pu-238 0.05%

Pu-239 0.01%

Pu-240 0.03%

Pu-241 0.42%

Am-241 0.74%

Am-243 0.02%

Cm-244 0.12%

Table 5-11. Nuclide mix information for 

West Valley uranium source term as of 

January 1, 2000. 

Nuclide Activity fraction

Ra-226 0.00%

Th-228 0.97%

Th-229 0.02%

Th-230 0.02%

U-232 0.93%

U-233 5.63%

U-234 61.87%

U-235 2.44%

U-236 8.88%

U-238 19.24%
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 74 of 152

Table 5-12. Nuclide mix information for West 

Valley plutonium source term as of January 1, 

2000. 

Nuclide Activity fraction α activity fraction

Pu-238 10.55% 44.65%

Pu-239 2.86% 12.10%

Pu-240 2.18% 9.23%

Pu-241 76.36% --

Am-241 8.04% 34.02%

The plutonium mixture in Table 5-12 is representative of aged material. Freshly separated plutonium 

during the operations era would not show appreciable 241Am content and would be best represented 

by the activity fractions given in Table 2-26. Table 5-13 summarizes the alpha activity fractions to use 

for fresh and aged plutonium mixtures at West Valley. The mixture for the fresh material is the 1971 

data from Table 2-26. Absorption type (M or S) should be selected based on what is known about the 

material involved in an intake. If the chemical form is unknown, then both type M and S should be 

used to see which results in the largest organ dose. The absorption type indicated by bioassay 

measurements should be used in cases where there are a sufficient number of measurements to 

make a determination. The absorption type selected for 241Am should be the same as for the 

corresponding plutonium matrix, i.e., either type M or type S. 

Table 5-13. Nuclide mix for freshly separated and aged 

plutonium at West Valley 

Nuclide

α activity fraction for plutonium mixtures

Freshly separated material Aged material

Pu-238 58.1% 44.65%

Pu-239 24.7% 12.10%

Pu-240 17.1% 9.23%

Pu-241 0.1% --

Pu-242 0.02% Not reported

Am-241 -- 34.02%

The same approach described above for selection of absorption type for plutonium should be applied 

for uranium and thorium. Absorption type should be selected based on which one yields the higher 

dose to the organ of interest unless there is sufficient bioassay data with which to make a 

determination. For uranium, users should select between absorption types F, M or S. For thorium 

users should select between absorption types M or S. 

The nuclide mix information for MFAP given in Table 5-10 is representative of aged material. Fresh 

material would show additional, shorter-lived nuclides. These are discussed in Section 2.7.1. 

Specific nuclide mix information that would aid in the evaluation of in vitro analysis results for gross 

beta or MFP is not available. Keely and Wenstrand (1971) describe an incident involving inhalation of 

MFAP by two workers. In vivo counting performed three days post-intake showed the activity to be 

predominantly 95Zr/Nb. The next largest contributor was 137Cs, followed by 106Ru/Rh and 60Co. 

However, these in vivo results do not represent the activity mix of the inhaled material or consider 

other contributors such as isotopes of strontium, yttrium, cerium, and promethium. 

The internal monitoring program at West Valley during operations was designed to detect chronic 

levels of intake corresponding to a threshold of 25% of applicable legal (10 C.F.R. pt. 20) limits. 

Radiation workers were included in routine in vitro and in vivo analysis programs, but the frequency 

and sensitivity of this monitoring was established relative to the 25% threshold for mandatory 

personnel monitoring. The routine in vivo monitoring program made use of the chest counter until 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 75 of 152

mid-1972. The counter was aligned to count 137Cs activity, which is cleared rapidly from the body 

following intake. The in vitro monitoring program for MFP would also have had limited sensitivity for 

non-chronic exposures. 

Users may employ ORAUT-OTIB-0002 (ORAUT 2007b) or ORAUT-OTIB-0018 (ORAUT 2005b) as 

appropriate for assigning internal dose for unmonitored workers or those for whom monitoring data 

are missing or incomplete. These documents should be used in accordance with their stated 

applicability and limitations. 

6.0

6.1 BASIS OF COMPARISON

EXTERNAL DOSIMETRY

From review of the available reference material, it appears external whole-body dose data for persons 

monitored at West Valley have always been reported in rem or fractions thereof. This includes dose 

to the skin of the whole body. However, the calibration methods used by the commercial dosimetry 

vendors or by the WVDP prior to becoming accredited under the DOE Laboratory Accreditation 

Program (DOELAP) are currently unknown. The WVDP external dosimetry program first became 

DOELAP accredited for whole body dose in January 1990 [16]. Users should therefore compute 

organ dose using conversion factors from exposure through 1989. From 1990 forward organ dose 

conversion factors from deep dose equivalent should be used. Skin dose measurements, if needed, 

should be used as reported. 

6.2 SITE HISTORICAL ADMINISTRATIVE PRACTICES

NFS used vendors to process its whole-body personnel monitors from the onset of monitoring 

throughout its tenure at the site. No definite date has been established for when personnel monitoring 

began. The transition of responsibility for the site from NFS to DOE under the West Valley 

Demonstration Act also saw a transition to on-site processing of personnel whole-body dosimetry. 

This practice continues to the present. A vendor is used for extremity monitoring. 

6.2.1

NFS Radiation Work Permits prescribe the following personnel dosimetry requirements (NFS 1973a, 

p. 2):

Site Access Requirements

• For plant entry: film badge

• For entry into Radiation Areas: film badge, dosimeters and current dose rate survey.

• For entry into High Radiation Areas: film badge, dosimeters and current dose rate survey.

NFS (1976) states all “employees and workers” entering radiation areas were required to wear film 

badges in addition to the indirect reading dosimeters. NFS (1974b, p. 4.0) states, “Each permanent

employee is issued a film badge and dosimeter ... at the Guard House as he enters [the] exclusion 

area and he is to wear it all times in the plant areas.” This is echoed in (Nelson 1973b, p. 7 of 

enclosure 2), which says, “First line control over the issuance of personnel monitoring devices is 

exercised at the Security Area Guard House.” Dosimeters and badges worn by people “normally 

assigned to radiation work” were to be returned to the badge racks after each shift (NFS 1974b, p. 

4.0). Temporary badges were supposed to be assigned if a worker lost his badge (Runion 1967, 

p. 2). 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 76 of 152

It appears the same dosimetry requirements applied to workers in the waste disposal areas. Workers 

in the area now known as the SDA wore “radiation badges and gamma dosimeters” (i.e., film badges 

and pocket meters, respectively) (Author unknown ca. 1968). 

NFS (1976) states neutron exposures were encountered in the FRS and the plutonium loadout area. 

“Neutron film badges” were therefore used in those areas. NFS (1974b, p. 4.0) states, “Special film 

badges sensitive to neutrons will be worn by personnel working in plutonium loadout and shipping 

areas when directed to do so. A satisfactory alternate procedure will be to place neutron films in 

typical locations and calculate exposures” using occupancy times. “On lower burnup fuels, Health 

and Safety will waive these requirements.” 

NFS’ practice was to assign extremity dosimeters (TLDs) when extremity dose rates were expected to 

exceed those for the skin of the whole body (Duckworth 1973). 

A memorandum titled “Personnel Accountability and Dosimetry” (Heacker 1972) discusses concerns 

about personnel accountability at West Valley during an accident scenario vis-à-vis their film badges. 

It states they could not be certain an individual had on the right badge or if they had a badge at all, 

and that the sign-in procedure was not rigorously followed. It mentions a “brown and white nuisance 

badge,” but says not all employees wore them. It also says badges were checked for contamination 

each shift at the guardhouse, but they could still have undetected contamination inside them. The 

impetus for these statements was to emphasize the need, in Heacker’s view, for photo badges such 

as those used at AEC sites. He used the photo badge in use at the time at Oak Ridge National 

Laboratory as an example. NFS’ records show isolated cases of personnel picking up the wrong 

dosimeter at the storage rack on their way in to the plant, but it does not appear this was a common 

occurrence or that it happened at a disproportionate frequency relative to other facilities where a 

similar number of people were badged. Once recognized, cases where individuals wore the wrong 

dosimeter were addressed by sending the badges for processing and assigning new badges to the 

affected individuals. After processing, the reported dose was assigned to the appropriate individual. 

Dose could also be assigned through the use of survey data or pocket chamber readings. 

Concurrence was sought from the affected individuals when such dose estimations were made. 

Under the current program, dosimetry is issued to radiation workers as follows, as “determined by the 

employee’s potential exposure” (Hoffman 1997):

• Radiation Worker I, quarterly exchange

• Radiation Worker II, quarterly exchange

• Radiation Worker I, monthly exchange

• Radiation Worker II, monthly exchange

Dosimetry is only issued to workers who meet the following criteria: 

• Have completed a baseline whole-body count or bioassay (or both)

• Are included on the periodic whole-body count or bioassay roster (or both)

• Have completed radiation worker training

• Have completed General Employee Training

• Have no radiological restrictions

• Have their radiation dose history on file or on request by the Radiation Protection department

Workers who are badged monthly are also issued direct-reading dosimeters, which along with the 

TLDs, are worn continuously by workers in Radiation Control Areas. In some cases, electronic 

dosimeters are used in addition to the TLDs and direct-reading dosimeters (Hoffman 1997). These 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 77 of 152

are used to supplement external dose information obtained from TLDs, and do not serve as a primary 

dosimeter. 

Visitor badges are issued to those having a “valid need” to enter areas requiring dosimetry. 

Supplemental dosimetry, if used, is placed at the location of the highest anticipated dose rate to 

monitor localized exposures (to extremities, for instance). 

During construction of the RHWF, all construction workers were trained as Radiation Workers and 

were issued a TLD while they were in the RHWF Radiological Buffer Area (RBA). Individuals who 

were expected to spend less than 80 hr/quarter in the RBA were classified as visitors and were not 

Radiation Worker trained or issued dosimetry. Some means of tracking the total time spent in the 

RHWF RBA was supposed to be in place and if a visitor reached the 80-hour limit, they could be 

subject to Radiation Worker training and issued dosimetry (Crotzer 2000, p. 2).

6.2.2

Personnel exposures at West Valley during plant operation were controlled relative to the legal 

maximums for 

Personnel Exposure Control

quarterly

• 3 rem to the whole body (or 5 rem/yr for “younger employees” (see Keely 1968a) 

exposure at that time: 

• 7.5 rem to the skin of the whole body

• 18.75 rem to the extremities

Personnel exposures were such that NFS had to utilize a substantial number of contract employees 

each quarter once the regular employees reached their limits. NFS used upwards of 1,000 contract 

employees per year, relative to a permanent staff of under 200. Controlling exposures to lower limits 

would have required substantial increases in personnel (e.g., see Wenstrand 1971e). 

In response to a number of extremity overexposures, NFS revised its SWPs in the third quarter of 

1967 to include limits for skin and extremity exposures (Lewis 1967b). Their Health and Safety 

Manual had been amended to include daily limits for skin and extremity exposure in April of that year 

(Keely 1967b). The individuals receiving extremity overexposures were maintenance or mechanical 

operators performing decontamination and maintenance on mechanical processing equipment such 

as manipulators. Lead gloves therefore procured for such tasks. Keely and Wenstrand (1971) says 

20,000-V lineman’s gloves were also used, though he does not indicate when this practice began. He 

also says if the gloves were too cumbersome to wear, workers changed their cotton gloves “every few 

minutes” to reduce the build up of contamination.

The individuals affected by the 1967 extremity overexposures were “removed from work which could 

give any further significant extremity dosage,” so it does not appear that they were removed from 

radiation work altogether (Lewis 1967b).

In a February 24, 1971, letter to its headquarters, NFS points out the majority of over-exposures at 

West Valley over its operating history were from chronic exposure and not acute events. In 

September 1969 they made efforts to reduce the rate of accumulated dose, but over-exposures still 

occurred. The practice was modified in the first quarter of 1970 by putting workers on dose rate 

restrictions once they exceeded 80% of their quarterly limit (Duckworth 1971). Thus, individuals were 

placed on dose rate restriction when they reached 2,400 mrem whole body, 6,000 mrem skin, or 

16,000 mrem extremity dose in an effort to “help control chronic over-exposures” (Wenstrand 1970). 

Exposure control practices were changed again in the second quarter of 1970 to where workers were 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 78 of 152

placed on daily dose restrictions once they exceeded 80% of their limit and total radiation work 

restriction when they reached 90% (Duckworth 1971).

As of May 1971, NFS controlled dose to contract employees to the following quarterly limits (Metzler 

1971): 

• 2 rem to the whole body

• 4 rem to the skin of the whole body

• 13 rem to the extremities

NFS began a practice of summing and recording dose results from pocket chambers and extremity 

TLDs for contract employees for comparison with these limits. The impetus was frequent occurrences 

of these individuals being called in for work whom were already on dose restrictions (Metzler 1971). 

NFS points out the lag time between collection and processing of film badges contributed to personnel 

exposure problems. They began sending their film badges to the processor by Special Delivery, 

which reduced the lag time to about 2 weeks (Duckworth 1971). In 1970, NFS was encouraged by its 

insurance carrier to find ways to reduce the delay time for the processing of film badges and to 

consider doing its own processing. Accumulated dose (for the purpose of determining individuals to 

place on work restriction) was based on film badge data when available and “dosimeter” (i.e., pocket 

chamber) data if not (Keely 1970a). A May 1972 insurance inspection report (Johnson & Higgins 

1972, p. 16) also mentions the long delay times for processing personnel dosimetry, saying the 

turnaround is “about three weeks from mailing to receipt of results.” This is said to be a decrease 

from what delays used to be (Johnson & Higgins 1972, p. 16).

6.2.3

6.2.3.1 Dose Recording Practices

Evaluation and Recording of Personnel Exposures

NFS determined skin and whole-body dose from film badges, and extremity dose from TLD finger 

dosimeters (NFS 1974b, p. 2.3). The film badge was the dosimeter of record. If the film had “an 

obviously incorrect exposure,” supplementary data, such as dosimeter readings and surveys were 

used to determine the exposure (NFS 1974b, p. 4.0). 

Per NFS, “All persons entering radiation areas or restricted areas identified in 10 CFR 20.202 are 

required to wear two indirect reading dosimeters with a range of 200 mR.” These dosimeters were 

read at the end of each shift and the reading recorded. These readings were totaled each week and 

transferred to the individual’s exposure card. The exposure card contained all of the information 

required on an AEC Form 5 (NFS 1976). Self-reading dosimeter results were supposed to have been 

recorded on the SWP if they exceeded 100 mR. A weekly dosimeter log was also kept (O’Reilly 

1975, p. 4 of inspection report). The indirect reading dosimeter totals provided a temporary value for 

whole-body exposure until film badge results were received (NFS 1976). 

Extremity TLDs were issued to those receiving exposures where skin dose was anticipated to exceed 

the whole-body dose. The extremity TLDs issued were to be returned to the Hot Lobby for processing 

upon completion of the work. Depending on what time work was completed, sometimes the TLDs 

were not read until the following morning (Duckworth 1973, p. 1). 

For dose to the skin of the whole body, the TLD results were used as a temporary measure until the 

film badge readings came back. For dose to the extremities, the TLD reading was the dose of record 

(NFS 1976). At one point NFS added a TLD disk to the film badge that could be read locally in an 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 79 of 152

effort to predict the film badge reading. They say the TLD disk and the film badge generally agreed to 

within 30%. 

Whole-body exposure data was received from the film badge supplier and recorded on individual’s 

exposure card. Exposure data from pencil dosimeter and TLD readouts was also compiled daily and 

recorded on an individual’s card. Lifetime totals were summarized quarterly (Duckworth 1972). 

Keely (1970b) suggests there were discrepancies between the “dosimeters” (pocket meters) and the 

film badges. “Overexposures” indicated by the pocket meters were not treated as true dose until the 

film badge result was available. 

Review of “Finger Dosimeter Records” forms from April 1972 showed entries under “Name” also 

included an indication of whether the dose was “skin” or “ext” (i.e., extremity) (NFS 1972b). 

The NFS records include several requests from the site asking the dosimetry vendors to change 

personnel dose results based on “dosimeter” (pocket chamber) readings (for examples, see Keely 

1968g,h). Both ask for personnel monitoring results to be changed based on “dosimeter” readings. 

The latter asks that a reported dose of 2,290 mR be changed to 165 mR. It is not known if the 

dosimetry vendors actually made these changes, and if so, where they would appear.

The requests from NFS for changes to film badge readings did not specify what prompted them. 

None of the requests seen had the corresponding dosimetry reports attached. An April 4, 1967, letter 

from Runion to the AEC says most rejections of film badges are due to overheating, accidental 

exposure to light, water damage, uneven exposure, or exposure through the back of the badge. It 

states if a badge is rejected for any reason, the person’s pocket dosimeter readings are summed and 

added to his previous film badge dose (Runion 1967b, p. 2). 

NFS made distinctions between NFS employees and non-NFS (e.g., Benz Construction workers) in its 

radiation exposure summaries. In the 1970 to 1971 era, Benz workers outnumbered NFS employees 

by approximately 5:1. NFS often gave detailed breakdowns of radiation exposure associated with 

specific job functions or tasks, but only for NFS employees. Data for Benz workers was all lumped 

together. They also appear to have distinguished between contract employees who worked at the 

reprocessing plant and those whom worked at the SDA (Nelson 1973b, p. 8 of enclosure). This 

reference states, “Contractor employee exposures are monitored by H&S [health and safety] by daily 

review of their dosimeter records.” 

Much of the external dose information in the NOCTS claim files for individuals whom worked at West 

Valley under NFS consists of a single sentence asserting total whole body, skin, and extremity dose 

for their period of employment. No individual monitoring data are provided. Dose reconstructors will 

therefore need to fractionate the reported total dose over the employment period a number of different 

ways so that latency is accounted for in the POC calculation in a manner most favorable to the 

claimant. This can be done, for example, by loading the reported total dose over the first few years of 

employment, the last few years, and evenly over the employment period and taking the result that 

yields the largest POC. However, in doing so the dose assigned for any individual year should not 

exceed the corresponding legal maximum. Annual legal maximums for the operations era are as 

follows (e.g., see Wenstrand 1971b). 

• 12 rem whole body

• 30 rem skin

• 75 rem extremity
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 80 of 152

6.2.3.2 Nuclear-Chicago Film Badge Exposure Reports

Nuclear-Chicago reported the “current reading” for an individual under either the heading “X-G-N” or 

“BETA,” with X-G-N referring to X-ray, gamma and neutron. X-ray plus gamma results and neutron 

results (if any) were reported separately under the “X-G-N” column, with the X-ray plus gamma result 

listed above the neutron value. Results in this column were in rem, with a vertical dashed line 

representing the decimal point. So for example, an entry of “123” to the right of the dashed line would 

be read as 123 mrem. In contrast, beta dose was reported to the nearest 10 mrem, so only two digits 

appeared to the right to the right of the vertical “decimal point” line. Hence, an entry of “25” to the 

right of the line would be read as 250 mrem. 

Nuclear-Chicago used an 8000 series number to identify neutron films under the “Film Number” 

column. Later Nuclear-Chicago exposure reports gave X-ray plus gamma and neutron results in 

separate columns. 

Later Nuclear-Chicago exposure reports indicated a badge type in the second column. These were 

as shown in Table 6-1. 

Table 6-1. Nuclear-Chicago film badge type codes. 

Type code Type

1 X-ray, beta, gamma clip-on

2 X-ray, beta, gamma wrist

3 X-ray, beta, gamma ring

5 Extended range X-ray, beta, gamma

14 X-ray, beta, gamma, and neutron clip-on

24 X-ray, beta, gamma, and neutron wrist

Frequency codes identified the exchange frequency. It appears a code of “1” indicated a monthly 

exchange and “5” indicated quarterly. 

On the older Nuclear-Chicago reports, values in columns labeled “H” indicated the total number of 

readings used to arrive at a given total. So for example, the value should be “3” at the end of a 

calendar quarter for a monthly subscriber. The “H” column only accommodated a single character, so 

the letters O, A, B, and C were used for 10, 11, 12, and 13, respectively. 

Entries in Column “D” indicated cases where badges could not be read at all or evaluated accurately. 

The codes shown in Table 6-2 were used. 

Table 6-2. Discrepancy codes and explanations. 

Discrepancy codes used in Column “D” 

of Nuclear-Chicago Exposure Reports Explanation

1 The film was not returned within 90 days. Evaluation was not 

possible due to aging of the sensitive film.

2 Film was apparently exposed to light. No reading was possible.

3 The film was blacked out, indicating either exposure to light or 

chemical fumes, or an exposure in excess of 600 R.

4 The film was soaked in water, over-heated, or otherwise 

mishandled in such a way that reading was not possible.

5 The film was exposed unevenly or partially obliterated, meaning 

the reading will reflect less information than usual, with a 

corresponding loss of accuracy.
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 81 of 152

Discrepancy codes used in Column “D” 

of Nuclear-Chicago Exposure Reports Explanation

6 Exposure occurred through the back of the badge, meaning a 

loss of accuracy due to self-shielding. 

Later Nuclear-Chicago reports include a column for skin dose. This was the sum of the X-ray plus 

gamma dose and any beta dose. 

6.2.3.3 Landauer Radiation Dosimetry Reports

The personnel exposure data in the Landauer reports are largely self-explanatory. Their reporting 

levels (ca. 1971) were 10 mrem for X-or gamma rays, 40 mrem for hard beta, 20 mrem for fast 

neutron, and 10 mrem for thermal neutron. Column 5 indicated the dosimeter type and Column 6 the 

exposure type. 

The dosimeter type codes typically seen for West Valley workers are “J” and “N,” corresponding to 

“Regular beta-gamma body film badge” and “Fast neutron-only body film badge.” See Attachment B 

for other dosimeter type codes that may be encountered. Table 6-3 lists the exposure type codes.

Table 6-3. Landauer exposure type codes. 

Code Exposure type

1 Total body

2 Skin dose

3 Right finger

4 Left finger

5 Right wrist

6 Left wrist

7 Other extremity

8 Other total body

Codes listed in column 4 for discrepancies or other notes are too numerous to include here. See 

Attachment B. 

6.3 SITE DOSIMETRY TECHNOLOGY

6.3.1

Whole-body monitoring was performed using film badges from the onset of personnel monitoring 

through August 1982. At least two film badge vendors were used: Nuclear-Chicago and Landauer. In 

October 1970 NFS discontinued film badge service with Nuclear-Chicago and ordered film badges 

from Eberline Instrument Corporation for a comparison evaluation with the Landauer badges (Keely 

1970c). However, it is not known if another vendor was actually used. 

Whole-Body Monitoring

The changeover from film badges to a Harshaw TLD system began in August 1982 (Daugherty 1986, 

p. 3). The Harshaw TLD system was replaced in September 1986 with a Panasonic TLD system, 

which is still in use. The external dosimetry program is presently accredited by the DOE Laboratory 

Accreditation Program (Hoffman 1997). 

6.3.2

NFS used TLD chips for extremity monitoring, which were read on site. These “finger dosimeters” 

went into service in March 1967 “on personnel exposed to high extremity dose rates” (Keely 1967a, p. 

Extremity Monitoring
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 82 of 152

1). A 1968 memorandum describes using TLD disks in black pouches attached with adhesive tape for 

extremity monitoring. It states they were good for single use in high exposure rate areas, but did not 

work well for continuous wearing by analytical laboratory workers (Keely 1968i). Wenstrand (1968d) 

says they are “using LiF TLD disks to monitor extremity exposure to laboratory analysts. The analyst 

is assigned two dosimeters and tapes at the beginning of each shift week and wears the dosimeters 

during the entire week. During the week, the analyst must change tapes on the dosimeters several 

times as the adhesive wears off.” NFS is requesting a “plastic ring dosimeter holder” to replace the 

“tapes.” 

A statistical evaluation was made of the extremity exposure of 76 work teams in the MSM Repair 

Shop during the fourth quarter of 1971. The 76 pairs of extremity dosimeters showed an average 

difference of 88% with a maximum of 275%. The minimum detectable dose for the extremity TLD was 

said to be 10 mrem (Duckworth 1973, p. 3). 

6.4 WORKPLACE RADIATION FIELDS

6.4.1

The West Valley reprocessing plant used the same solvent extraction process (PUREX) used at AEC 

plutonium production facilities, notably, the Hanford site. The workplace photon energy spectra, 

therefore, would likewise have been dominated by high-energy photons from mixed fission and 

activation products, with a low-energy component representing down-scatter. This is shown in the 

photon energy distribution information provided in the Hanford site profile for the separation plants. 

The photon energy selections established for the separations plants are 75% >250 keV and 25% 

30 to 250 keV (ORAUT 2006c). The same selections should be used for photon exposures in the 

reprocessing plant at West Valley. Likewise, beta (electron) exposures should be treated as 100% > 

15 keV. These selections should be applied for all facilities and areas over the entire time period of 

concern (1965 – present). 

Beta and Photon

Workers in the waste burial areas received exposures from handling and emplacement of waste 

containers and items. These included waste materials and items from commercial entities (in the 

case of the SDA) and HLW from the reprocessing plant. LLW from the reprocessing plant were also 

disposed of at the SDA. Users should assume the energy groups for photons and electrons as given 

above for the processing plant [17]. 

6.4.2

Neutrons could have been encountered in the FRS (cask handling operations), the Product Packaging 

and Handling (PPH), and the Product Packaging and Shipping (PPS) areas from spontaneous fission 

sources and (α, n) reactions. Given a lack of information on workplace neuron spectra at West 

Valley, users should assume 100% of any neutron dose was from the 0.1 to 2.0 MeV energy group

[18]. 

Neutron

Any positive neutron dose reported for West Valley workers based on personnel dosimetry or survey 

data should be doubled to reflect ICRP Publication 60 radiation weighting factors (ICRP 1991). This 

would apply to any missed neutron dose assigned as well. 

6.4.3

Table 6-4 summarizes the Interactive RadioEpidemiological Program (IREP) energy group selections 

that should be made for West Valley cases.

Summary
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Table 6-4. Selection of radiation energies for West Valley facilities. 

Area Radiation type Energy selection Fraction

All areas Photon 30–250 keV 25%

>250 keV 75%

All areas Electron >15 keV 100%

FRS, PPH, PPS Neutron 0.1–2.0 MeV 100%

6.5 MISSED DOSE

Exchange frequencies, minimum detectable levels (MDLs) and maximum annual missed dose are 

given in the tables below for the personnel dosimetry badges used at West Valley. Table 6-5 provides 

data for deep dose. Table 6-6 is for nonpenetrating radiation. The MDLs for deep and 

nonpenetrating dose for the commercial film badge reflect the guidance provided in (ORAUT 2006d) 

and (ORAUT 2005c), respectively. The MDLs for the Harshaw and Panasonic TLDs are given in 

DOE-era dosimetry reports in NOCTS claim files. 

An asserted MDL for the commercial film badges used at West Valley has not been identified. 

Landauer’s reporting limit for its film badge was 40 mrem for hard beta. 

NFS asserted the MDL for its extremity TLD chips was 10 mrem. It is assumed this value is for 

penetrating radiation. Slawson (1999) reports the MDL for the Landauer Type U extremity dosimeter 

ranges from 5.1 to 5.9 mrem and averages 5.5 mrem. Claim files in NOCTS give a MDL for the 

Table 6-5. Exchange frequencies, MDLs, and maximum annual missed deep dose for personnel 

dosimeters. 

Period Dosimeter

MDL 

(rem)

Exchange

frequency

Maximum annual 

missed dose (rem)

Until 1982 Commercial film badge 0.040 Weekly (n = 52) 1.040

Monthly (n = 12) 0.240

1982–August 1986 Harshaw TLD 0.015 Monthly (n = 12) 0.090

Quarterly (n = 4) 0.030

September 1986–present Panasonic TLD 0.010 Monthly (n = 12) 0.060

Quarterly (n = 4) 0.020

Table 6-6. Exchange frequencies, MDLs, and maximum annual missed nonpenetrating dose for 

personnel dosimeters. 

Period Dosimeter

MDL 

(rem)

Exchange 

frequency

Maximum annual 

missed dose (rem)

Until August 1982 Commercial film badge 0.050 Weekly (n = 52) 1.300

Monthly (n = 12) 0.300

August 1982–August 1986 Harshaw TLD 0.030 Monthly (n = 12) 0.180

Quarterly (n = 4) 0.060

September 1986–present Panasonic TLD 0.030 Monthly (n = 12) 0.180

Quarterly (n = 4) 0.060

Landauer Type U finger ring of 30 mrem/exchange. It is assumed the difference reflects that between 

penetrating and nonpenetrating radiation. The extremity dosimeters were used on an ad hoc basis, 

so no annual missed dose can be asserted. 

Practices for assigning personnel neutron dosimetry at West Valley during operations have not been 

identified. If necessary, users should assume a neutron MDL of 80 mrem, consistent with that 

asserted for neutron film dosimeters in previous site profiles (e.g., Hanford). Any neutron missed 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 84 of 152

dose assigned using this MDL should be doubled to account for ICRP Publication 60 radiation 

weighting factors (ICRP 1991). 

6.6 UNCERTAINTY

Film badges normally over-respond in the workplace except for where plutonium or americium are the 

primary contributors to external dose (ORAUT 2006d). Dose contributions from low-energy photons 

are not believed to ever have been prevalent at West Valley given the predominance of photon 

exposures from the ventilation and acid recovery systems. Use of recorded film badge dose for 

photon exposure is therefore believed to be favorable to the claimant. 

Specific information pertaining to the uncertainty associated with recorded personnel dose at West 

Valley has not been identified. Uncertainties to assign for positive recorded dose are given in Table 

6-7. These have been reproduced from information provided in the external dosimetry technical basis 

documents for the Y-12 and X-10 sites. Additional discussion of the uncertainties in film badge data 

may be found in (Morgan 1961). 

Table 6-7. Uncertainties for positive recorded dose at West Valley. 

Dosimeter Period of use Photon Beta Neutron

Commercial film badge 1965–1982 ±30% ±30% ±50%

Commercial TLD 1982–present ±15% ±15% ±15%

7.0

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

information, conclusions, and recommendations to assist in the process of worker dose 

reconstruction. These callouts are listed in this section with information that identifies the source and 

justification for each item. Conventional references are provided in the next section that link data,

quotations, and other information to documents available for review on the Oak Ridge Associated 

Universities (ORAU) Team servers.

ATTRIBUTIONS AND ANNOTATIONS

This site profile was authored by Robert E. Burns, Jr., Certified Health Physicist (CHP). Many of the 

following attributions, therefore, represent conclusions and impressions drawn from detailed review of 

records from when the West Valley site was operating. 

[1] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

Elevated exposure rates existed throughout the process building and exterior to it during the 

operations era. Review of operating records showed numerous instances of removable 

contamination being spread throughout multiple areas of the process building. Contamination 

incidents sometimes would not be recognized until it reached radiological buffer areas at 

personnel ingress/egress points. 

[2] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

This is apparent through review of plant records and AEC inspection reports. The AEC often 

remarked about the radiation protection program being implemented vis-a-vis legal 

maximums. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 85 of 152

[3] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

Elevated exposure rates from acid recovery system components and contamination from 

major spills affected all areas of the plant. Plant records indicate persons not normally 

involved in radiation work were sometimes brought in to support tasks in the process building 

when personnel exposure limits became a constraint. Table 2-7 shows acid recovery 

supervisors collectively averaged 6 person-rem/yr. between 1970 and 1971. The work area 

for these individuals was “offices.” 

[4] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The NDA is within the DOE fence line, and the WVDP built a Liquid Pre-Treatment System on 

the northeast corner of the NDA ca. 2004. The purpose of the system was to remove organics 

(solvent) from groundwater around the NDA.

[5] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The information on the contact dose rate from the unshielded glass logs and the impact of 

elevated exposure rates on release surveys is from discussions the author had with site 

radiation protection manager during a site visit in June 2006. 

[6] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

These three streams represent the three principal processes when the plant was operating: 

mechanical and chemical preparation of the fuel, separation of the uranium and plutonium 

products, and handling of the associated chemical wastes. 

[7] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The fuels processed at West Valley consisted of both metal and oxide forms. 

[8] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The plutonium product from the PUREX process is in an aqueous nitrate form. Plutonium 

contamination will undergo oxidation as it ages, reducing its solubility. 

[9] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The PUREX process separates the americium present in the feed stream from the plutonium 

product stream. 241Am slowly grows back in over time from decay of 241Pu. 

[10] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The fuels processed at West Valley consisted of both metal and oxide forms.

[11] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The nuclides listed represent those observed in in vivo counting data for operations era 

workers. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 86 of 152

[12] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

This is apparent from review of NFS documents discussing atmospheric releases and 

correspondence between NFS and the AEC. 

[13] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

No references have been identified suggesting NFS ever changed its site access requirements 

for personnel monitoring for external exposure. 

[14] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

170 mrem/yr was the administrative limit for annual dose to unmonitored workers as of 1986 

(Roberts 1986). To be favorable to the claimant, it is being assumed unmonitored individuals 

received an annual dose equal to this limit for the period 1982 – 1993. Likewise, for 1994 

forward an annual dose equal to the 100 mrem/yr limit of 10 CFR 835 is being assumed. 

[15] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

(Wenstrand 1968c) is a letter from NFS to Eberline requesting information on their ability to 

support a fecal monitoring program. Plant records show a routine fecal analysis program was 

in place as of 1970. 

[16] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

Communication between Bob Burns and Paul Szalinski, CHP, Integrated Environmental 

Management, on March 15, 2007. 

[17] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

The photon energy spectra at the waste burial sites are not expected to differ from those in the 

processing plant in a way that would not be favorable to the claimant. 

[18] Burns, Jr. Robert E., CHP. Shonka Research Associates. Senior Health Physicist. March 

2007. 

This neutron energy group is representative of fission neutron spectra and is also most 

favorable to the claimant. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 87 of 152

REFERENCES

ASDA (New York State Atomic and Space Development Authority) and NFS (Nuclear Fuel Services), 

1971, Plutonium Storage Agreement between New York State Atomic and Space 

Development Authority and Nuclear Fuel Services, May 3. [SRDB Ref ID: 28316, p. 205]

Author unknown, ca. 1967, "Explanation of Exposure Report Data." [SRDB Ref ID: 24431, p. 222]

Author unknown, ca. 1968, "Facilities Description." [SRDB Ref ID: 29117, p. 237]

Author unknown, 1968, "Radiation-Contamination Status Report, May 1968," memorandum to R. B. 

Keely, West Valley, New York, June 3. [SRDB Ref ID: 24422, p. 178]

Author unknown, 1969, "Head End Vent., P.S.A.R., Third revision pre. Subm.," January 15. [SRDB 

Ref ID: 29127]

Author unknown, 1974, results of radiation counts. [SRDB Ref ID: 24404, p. 1780] 

Author unknown, ca. 1974, "Site Boundary Concentrations of α and β Releases from FRP Stack 

Annual β Sector Width Averaged."

Bailey, W. J., 1990, West Valley Facility Spent Fuel Handling, Storage, and Shipping Experience, 

Pacific Northwest National Laboratory, Richland, Washington, November. [SRDB Ref ID: 

30229] 

Birchler, J. E., 1970, "Plant Cleanout Between Processing of a Thorium Fuel and a Low Enriched 

Uranium Fuel," NFS-PS-138, memorandum to J. R. Clark, West Valley, New York, December 

16. [SRDB Ref ID: 26331, p. 33]

Clark, J. R., 1968a, "USAEC Compliance Inspection," memorandum to Dr. E. D. North, Nuclear Fuel 

Services, July 24. [SRDB Ref ID: 28343, p. 281]

Clark, J. R., 1968b, "Compliance with AEC Letter of December 28, 1968," memorandum to Dr. E. D. 

North, Nuclear Fuel Services, April 8. [SRDB Ref ID: 24455, p. 87]

Clark, J. R., 1968c, untitled memorandum to E. D. North, January 26.

Clark, J. R., 1968d, "Compliance Status," memorandum to Dr. E. D. North, Nuclear Fuel Services, 

May 21. [SRDB Ref ID: 28343, p. 317]

Clark, J. R., 1968e, "Criteria for Removal of Dog Filters," memorandum to Distribution, April 8. [SRDB 

Ref ID: 28343, p. 328]

Clark, J. R., 1979, untitled memorandum to C. B. Woodhall, November 5. 

Crotzer, M. E., 2000, "Thermoluminescent Dosimeter (TLD) Badge Requirements During Remote 

Handled Waste Facility Construction," memorandum to D. K. Pletz, IO:00:0036, West Valley 

Nuclear Services Company, West Valley, New York, July 6. [SRDB Ref ID: 25642]

Daugherty, H. F., 1986, "Handling and Disposition of Radioactively Contaminated Soil," memorandum 

to J. L. Knabenschuh, FA:86:0097, West Valley, New York, August 28. [SRDB Ref ID: 25992]
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DOE (U.S. Department of Energy), 1999, Plutonium and Uranium Recovery from Spent Fuel 

Reprocessing by Nuclear Fuel Services at West Valley, New York from 1966 to 1972, 

Revised, Office of Defense Programs, Washington, D.C., November. [SRDB Ref ID: 6251, p. 

25]

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Wenstrand, T. K., 1969a, "Radiation - Contamination Status Report, September 1969," memorandum 

to R. B. Keely, Nuclear Fuel Services, West Valley, New York, October 6. [SRDB Ref ID: 

24422, p. 135]

Wenstrand, T. K., 1969b, "Radiation Contamination Status Report, August 1969," memorandum to J. 

P. Duckworth, West Valley, New York, September 3. [SRDB Ref ID: 24422, p. 137]

Wenstrand, T. K., 1969c, "Radiation-Contamination Status Report - October 1969," memorandum to 

R. B. Keely, West Valley, New York, November 6. [SRDB Ref ID: 24422, p. 132]

Wenstrand, T. K., 1970, "Radiation Work Restrictions," memorandum to distribution, West Valley, 

New York, April 10. 

Wenstrand, T. K., 1971a, "Health and Safety Monthly Report - October 1971," memorandum to J. P. 

Duckworth, West Valley, New York, November 5. [SRDB Ref ID: 24422, p. 56]

Wenstrand, T. K., 1971b, "Summary of 1970 Radiation Exposures," memorandum to J. P. Duckworth, 

West Valley, New York, March 9. [SRDB Ref ID: 29089, p. 134]

Wenstrand, T. K., 1971c, "Health and Safety Monthly Report - November 1971," memorandum to J. 

P. Duckworth, West Valley, New York, December 8. [SRDB Ref ID: 24422, p. 54]

Wenstrand, T. K., 1971d, "Health and Safety Monthly Report - May 1971," memorandum to J. P. 

Duckworth, West Valley, New York, June 7. [SRDB Ref ID: 24422, p. 64]

Wenstrand, T. K., 1971e, "Reduction of Whole Body Exposures to 5 Rems/Year," memorandum to J. 

P. Duckworth, West Valley, New York, July 1. 

Wenstrand, T. K., 1971f, "Safety and Industrial Relations Monthly Report - April 1971," memorandum 

to J. P. Duckworth, West Valley, New York, May 7. [SRDB Ref ID: 24422, p. 66]

Wenstrand, T. K., 1971g, "Contamination Around HEV," memorandum to B. E. Knight, West Valley, 

New York, October 12. [SRDB Ref ID: 24437, p. 1148]

Wenstrand, T. K., 1972a, Plant Radiation Status March 15, 1972, memorandum to J. P. Duckworth, 

West Valley, New York, March 17. [SRDB Ref ID: 28325, p. 12]

Wenstrand, T. K., 1972b, "Health and Safety Monthly Report - January 1972," memorandum to J. P. 

Duckworth, West Valley, New York, February 8. [SRDB Ref ID: 24422, p. 50]
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 98 of 152

Wenstrand, T. K., 1972c, "Radiation Exposure for 1971," memorandum to J. P. Duckworth, June 1. 

[SRDB Ref ID: 29089, p. 118]

Wenstrand, T. K., 1972d, "Status of Plant Decontamination," memorandum to J. P. Duckworth, West 

Valley, New York, July 27. [SRDB Ref ID: 28325, p. 2]

Wenstrand, T. K., 1972e, Plutonium Nitrate Contaminated Puncture Wound, West Valley, New York. 

Wenstrand, T. K., 1972f, "Health and Safety Monthly Report - February 1972," memorandum to J. P. 

Duckworth, West Valley, New York, March 8. [SRDB Ref ID: 24422, p. 48]

Wenstrand, T. K., 1972g, "Health and Safety Monthly Report - May, 1972," memorandum to J. P. 

Duckworth, West Valley, New York, June 7. [SRDB Ref ID: 24422, p. 42] 

Wenstrand, T. K., 1972h, "Health and Safety Monthly Report - December 1971," memorandum to J. 

P. Duckworth, West Valley, New York, January 2. [SRDB Ref ID: 24422, p. 52]

Wenstrand, T. K., 1972i, "Health & Safety Monthly Report - June 1972," memorandum to J. P. 

Duckworth, West Valley, New York, July 10. [SRDB Ref ID: 24422, p. 41]

Wenstrand, T. K., 1972j, "Health and Safety Monthly Report - April 1972," memorandum to J. P. 

Duckworth, West Valley, New York, May 11. [SRDB Ref ID: 24422, p. 44]

Wenstrand, T. K., 1972k, "Health and Safety Monthly Report - March 1972," memorandum to J. P. 

Duckworth, West Valley, New York, April 13. [SRDB Ref ID: 24422, p. 46]

Wenstrand, T. K., 1973, "Radiation Levels at WTF Shelter," memorandum to B. E. Knight, West 

Valley, New York, April 19. [SRDB Ref ID: 29089, p. 67]

Wenstrand, T. K., 1974, "LWA and UWA Pump Changes," memorandum to J. R. Clark, West Valley, 

New York, December 5. [SRDB Ref ID: 28398, p. 10]

WVDP (West Valley Demonstration Project), 2005, "Summary of Facility Radiological Status by Area," 

Appendix C, WVDP Decommissioning Plan, Rev. 0, Draft 2. [SRDB Ref ID: 30280] 

WVNSC (West Valley Nuclear Services Company), 1983, 1982 Environmental Monitoring Program 

Report For The West Valley Demonstration Project, West Valley, New York, May. 

WVNSC (West Valley Nuclear Services Company), 1986, 1985 Environmental Monitoring Program 

Report For The West Valley Demonstration Project, WVDP-040, West Valley, New York, 

March. 

WVNSC (West Valley Nuclear Services Company), 1988, Personnel In Vivo

WVNSC (West Valley Nuclear Services Company), 1990a, 

Counting, RC-DOS-6, 

Rev. 0, West Valley Demonstration Project, West Valley, New York, November 2. [SRDB Ref 

ID: 24207]

In Vitro

WVNSC (West Valley Nuclear Services Company), 1990b, 

Monitoring, RC-DOS-7, Rev. 1, 

West Valley Demonstration Project, West Valley, New York, July 31. [SRDB Ref ID: 24193]

In Vitro Monitoring, RC-DOS-7, Rev. 2, 

West Valley Demonstration Project, West Valley, New York, December 7. [SRDB Ref ID: 

24194]
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 99 of 152

WVNSC (West Valley Nuclear Services Company), 1990c, Interpretation of In Vitro

WVNSC (West Valley Nuclear Services Company), 1993, Personnel 

Bioassay 

Measurements, RC-DOS-30, Rev. 0, West Valley Demonstration Project, West Valley, New 

York, October 1. [SRDB Ref ID: 24207]

In Vivo

WVNSC (West Valley Nuclear Services Company), 1994, Sealed Rooms Paper Characterization, 

WVDP-RFI-016, Rev. 0, West Valley, New York, May. [SRDB Ref ID: 30284] 

Counting, RC-DOS-6, 

Rev. 2, West Valley Demonstration Project, West Valley, New York, May 12. [SRDB Ref ID: 

24186] 

WVNSC (West Valley Nuclear Services Company), 1997, Interpretation of Bioassay Measurements, 

RC-DOS-30, Rev. 2, West Valley Demonstration Project, West Valley, New York, September 

30. [SRDB Ref ID: 24209] 

WVNSC (West Valley Nuclear Services Company), 1999, Area and Emergency Dosimetry Badges, 

RC-DOS-34, Rev. 6, West Valley Demonstration Project, West Valley, New York, October 29. 
Document No. ORAUT-TKBS-0057 Revision No. 00 Effective Date: 06/22/2007 Page 100 of 152

TABLE OF CONTENTS

SECTION TITLE

A1.0 Acid Recovery Cell...............................................................................................................

PAGE

106

A1.1 Radiological Conditions ............................................................................................106

A1.1.1 External Exposure Data ............................................................................. 106

A2.0 Acid Recovery Pump Room ................................................................................................. 106

A2.1 Radiological Conditions ............................................................................................107

A2.1.1 External Exposure Data ............................................................................. 107

A2.1.2 Removable Contamination Data................................................................. 107

A3.0 Analytical Aisle, Laboratories, and North Operating Aisle..................................................... 107

A3.1 Radiological Conditions ............................................................................................108

A3.1.1 External Exposure Data ............................................................................. 108

A3.1.2 Removable Contamination Data................................................................. 109

A4.0 Cell Access Aisle ................................................................................................................. 109

A4.1 Radiological Conditions ............................................................................................109

A4.1.1 Removable Contamination Data................................................................. 109

A4.1.2 Airborne Contamination Data ..................................................................... 109

A5.0 Cell Crane Rooms................................................................................................................ 109

A5.1 Impact of the Thorex Campaign on Airborne Activity Levels in the Cell 

Crane Rooms ...........................................................................................................110

A5.2 General Purpose Cell Crane Room and Extension................................................... 110

A5.2.1 Radiological Conditions.............................................................................. 110

A5.2.2 External Exposure Data ............................................................................. 110

A5.2.3 Removable Contamination Data................................................................. 111

A5.2.4 Airborne Contamination Data ..................................................................... 111

A5.3 Process Mechanical Cell Crane Room ..................................................................... 111

A5.3.1 External Exposure Data ............................................................................. 111

A5.3.2 Removable Contamination Data................................................................. 111

A5.3.3 Airborne Contamination Data ..................................................................... 111

A5.4 Chemical Process Cell Crane Room......................................................................... 112

A5.4.1 External Exposure Data ............................................................................. 112

A5.4.2 Removable Contamination Data................................................................. 112

A5.4.3 Airborne Contamination Data ..................................................................... 112

A6.0 Chemical Operating Aisle and Chemical Viewing Aisle ........................................................ 112

A6.1 Radiological Conditions ............................................................................................112

A6.1.1 External Exposure Data ............................................................................. 112

A6.1.2 Removable Contamination Data................................................................. 113

A7.0 Chemical Process Cell ......................................................................................................... 113

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A7.1 Radiological Conditions ............................................................................................113

A8.0 Control Room....................................................................................................................... 113

A8.1 Radiological Conditions ............................................................................................113

A8.1.1 External Exposure Data ............................................................................. 113

A9.0 Equipment Decontamination Room and Equipment Decontamination Room 

Viewing Aisle........................................................................................................................ 114

A9.1 Radiological Conditions ............................................................................................114

A9.1.1 External Exposure Data ............................................................................. 114

A9.1.2 Removable Contamination Data................................................................. 114

A9.1.3 Airborne Contamination Data ..................................................................... 114

A10.0 Extraction and Product Purification Cells.............................................................................. 114

A10.1 Radiological Conditions ............................................................................................115

A10.1.1 External Exposure Data ............................................................................. 115

A10.1.2 Removable Contamination Data................................................................. 115

A10.1.3 Airborne Contamination Data ..................................................................... 116

A11.0 Extraction Chemical Room................................................................................................... 116

A11.1 Radiological Conditions ............................................................................................116

A11.1.1 External Exposure Data ............................................................................. 116

A11.1.2 Removable Contamination Data................................................................. 117

A12.0 Extraction Sample Aisle and Process Sample Cell 2............................................................ 117

A12.1 Radiological Conditions ............................................................................................117

A12.1.1 Removable Contamination Data................................................................. 117

A12.1.2 Airborne Contamination Data ..................................................................... 117

A13.0 Fuel Receiving and Storage................................................................................................. 117

A13.1.1 Radiological Conditions.............................................................................. 118

A13.1.2 Pool Water Concentration Data.................................................................. 119

A13.1.3 External Exposure Data ............................................................................. 120

A13.1.4 Removable Contamination Data................................................................. 121

A13.1.5 Airborne Contamination Data ..................................................................... 122

A14.0 Guard House........................................................................................................................ 122

A14.1 Radiological Conditions ............................................................................................122

A15.0 General Operating Aisle....................................................................................................... 123

A15.1 Radiological Conditions ............................................................................................123

A15.1.1 Exposure Rate Data................................................................................... 123

A15.1.2 Airborne Contamination Data ..................................................................... 123

A16.0 General Purpose Cell........................................................................................................... 123

A16.1 Radiological Conditions ............................................................................................123

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A16.1.1 External Exposure Data ............................................................................. 124

A16.1.2 Removable Contamination Data................................................................. 124

A17.0 Head End Ventilation Building and Crane Room .................................................................. 124

A17.1 Radiological Conditions ............................................................................................124

A17.1.1 External Exposure Data ............................................................................. 124

A17.1.2 Removable Contamination Data................................................................. 124

A18.0 Hot Acid Cell ........................................................................................................................ 125

A18.1 Radiological Conditions ............................................................................................125

A18.1.1 External Exposure Data ............................................................................. 125

A18.1.2 Removable Survey Data ............................................................................ 125

A19.0 Hot Shop.............................................................................................................................. 125

A19.1 Radiological Conditions ............................................................................................125

A19.1.1 External Exposure Data ............................................................................. 125

A19.1.2 Removable Contamination Data................................................................. 125

A20.0 Instrument Shop/Old Instrument Shop ................................................................................. 126

A20.1 Radiological Conditions ............................................................................................126

A20.1.1 External Exposure Data ............................................................................. 126

A20.1.2 Removable Contamination Levels.............................................................. 126

A21.0 Interceptor and New Interceptor........................................................................................... 126

A21.1 Radiological Conditions ............................................................................................127

A21.1.1 External Exposure Data ............................................................................. 127

A22.0 Laundry................................................................................................................................ 127

A22.1 Radiological Conditions ............................................................................................127

A22.1.1 External Exposure Data ............................................................................. 127

A22.1.2 Removable Contamination Data................................................................. 127

A22.1.3 Airborne Contamination Data ..................................................................... 127

A23.0 Load-in Facility..................................................................................................................... 128

A24.0 Low Level Waste Treatment Facility..................................................................................... 128

A24.1 Radiological Conditions ............................................................................................128

A24.1.1 External Exposure Data ............................................................................. 128

A25.0 Low Level Wastewater Treatment Building .......................................................................... 128

A25.1 Radiological Conditions ............................................................................................129

A25.1.1 External Exposure Data ............................................................................. 129

A25.1.2 Removable Contamination Data................................................................. 129

A26.0 Lower Extraction Aisle.......................................................................................................... 129

A26.1 Radiological Conditions ............................................................................................129

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A26.1.1 External Exposure Data ............................................................................. 129

A27.0 Lower Warm Aisle................................................................................................................ 129

A27.1 Radiological Conditions ............................................................................................130

A27.1.1 External Exposure Data ............................................................................. 130

A27.1.2 Removable Contamination Data................................................................. 130

A27.1.3 Airborne Contamination Data ..................................................................... 130

A28.0 Main Office Building ............................................................................................................. 131

A28.1 Radiological Conditions ............................................................................................131

A28.1.1 External Exposure Data ............................................................................. 131

A28.1.2 Removable Contamination Data................................................................. 131

A29.0 Maintenance Shop and Warehouse ..................................................................................... 131

A29.1 Radiological Conditions ............................................................................................132

A30.0 Manipulator Repair Room .................................................................................................... 132

A30.1 Radiological Conditions ............................................................................................132

A30.1.1 External Exposure Data ............................................................................. 132

A30.1.2 Removable Contamination Data................................................................. 132

A31.0 Mechanical Operating Aisle (North, East and West)............................................................. 132

A31.1 Radiological Conditions ............................................................................................133

A31.1.1 External Exposure Data ............................................................................. 133

A31.1.2 Removable Contamination Data................................................................. 133

A31.1.3 Airborne Contamination Data ..................................................................... 134

A32.0 MSM Shop/Contact Size Reduction Facility ......................................................................... 134

A32.1 Radiological Conditions ............................................................................................134

A32.1.1 External Exposure Data ............................................................................. 134

A32.1.2 Removable Contamination Data................................................................. 134

A33.0 Off Gas Aisle and Process Sample Cell 3 ............................................................................ 135

A33.1 Radiological Conditions ............................................................................................135

A33.1.1 External Exposure Data ............................................................................. 135

A33.1.2 Removable Contamination Data................................................................. 135

A33.1.3 Airborne Contamination Data ..................................................................... 136

A34.0 Off Gas Blower Room .......................................................................................................... 136

A34.1 Radiological Conditions ............................................................................................136

A34.1.1 External Exposure Data ............................................................................. 136

A34.1.2 Removable Contamination Data................................................................. 137

A35.0 Off Gas Cell ......................................................................................................................... 137

A35.1 Radiological Conditions ............................................................................................137

A35.1.1 External Exposure Data ............................................................................. 137

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A36.0 Permanent Ventilation System Building................................................................................ 138

A36.1 Radiological Conditions ............................................................................................138

A36.1.1 External Exposure Data ............................................................................. 138

A36.1.2 Removable Contamination Data................................................................. 138

A37.0 Process Chemical Room...................................................................................................... 138

A37.1 Radiological Conditions ............................................................................................138

A37.1.1 External Exposure Data ............................................................................. 138

A37.1.2 Removable Contamination Data................................................................. 138

A38.0 Process Mechanical Cell...................................................................................................... 139

A38.1 Radiological Conditions ............................................................................................139

A39.0 Product Packaging and Shipping, Product Packaging and Handling, Uranium 

Product Cell, and Uranium Loadout Areas ........................................................................... 139

A39.1 Radiological Conditions ............................................................................................140

A39.1.1 External Exposure Data ............................................................................. 140

A39.1.2 Removable Contamination Data................................................................. 140

A39.1.3 Airborne Contamination Data ..................................................................... 140

A40.0 Pulser Equipment Aisle ........................................................................................................ 140

A40.1 Radiological Conditions ............................................................................................141

A40.1.1 Airborne Contamination Data ..................................................................... 141

A41.0 Remote Handled Waste Facility ........................................................................................... 141

A42.0 Sample Storage Cell, Analytical Hot Cells, and Analytical Decontamination 

Aisle..................................................................................................................................... 141

A43.0 Scrap Removal Area............................................................................................................ 141

A43.1 Radiological Conditions ............................................................................................142

A43.1.1 External Exposure Data ............................................................................. 142

A43.1.2 Removable Contamination Data................................................................. 142

A43.1.3 Airborne Contamination Data ..................................................................... 142

A44.0 Solvent Storage Terrace ...................................................................................................... 142

A44.1 Radiological Conditions ............................................................................................142

A44.1.1 External Exposure Data ............................................................................. 142

A44.1.2 Removable Contamination Data................................................................. 143

A45.0 Supernatant Treatment System Support Building................................................................. 143

A45.1 Radiological Conditions ............................................................................................143

A45.1.1 External Exposure Data ............................................................................. 143

A45.1.2 Removable Contamination Data................................................................. 143

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A46.0 Upper Extraction Aisle.......................................................................................................... 143

A46.1 Radiological Conditions ............................................................................................143

A46.1.1 External Exposure Data ............................................................................. 143

A46.1.2 Removable Contamination Data................................................................. 144

A47.0 Upper Warm Aisle and Process Sample Cell 1 .................................................................... 144

A47.1 Radiological Conditions ............................................................................................144

A47.1.1 External Exposure Data ............................................................................. 144

A47.1.2 Removable Contamination Data................................................................. 145

A47.1.3 Airborne Contamination Levels .................................................................. 146

A48.0 Utility Room and Stairwells................................................................................................... 146

A48.1 Radiological Conditions ............................................................................................146

A48.1.1 External Exposure Data ............................................................................. 146

A48.1.2 Removable Contamination Data................................................................. 147

A48.1.3 Airborne Concentration Data...................................................................... 148

A49.0 Ventilation Exhaust Cell ....................................................................................................... 148

A49.1 Radiological Conditions ............................................................................................148

A49.1.1 External Exposure Data ............................................................................. 148

A50.0 Ventilation Systems.............................................................................................................. 148

A50.1 Main Ventilation System ...........................................................................................148

A50.2 Head End Ventilation System ...................................................................................150

A51.0 Ventilation Wash Room........................................................................................................ 150

A51.1 Radiological Conditions ............................................................................................150

A51.1.1 External Exposure Data ............................................................................. 151

A51.1.2 Airborne Contamination Data ..................................................................... 151

A52.0 Waste Tank Farm................................................................................................................. 151

A52.1 Radiological Conditions ............................................................................................151

A52.1.1 External Exposure Data ............................................................................. 151

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DESCRIPTIONS OF OPERATIONS AREAS AND SYSTEMS WITH EMPHASIS 

ON RADIOLOGICAL CONDITIONS

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This section describes specific facilities and areas where personnel exposures occurred, with an 

emphasis on radiological conditions and occurrences. 

The ground level of the reprocessing plant was the 100-ft elevation. There was one level below the 

ground level and four levels above. Attachment B contains drawings of the reprocessing plant and 

the site. 

A1.0 ACID RECOVERY CELL

The Acid Recovery Cell (ARC) was in the southwest corner of the plant on the 114.5-ft level. It 

housed the fractionator and other equipment used for recovery of process acid. The recovered acid 

would be mixed with chemicals in the XCR before reuse. Significant dose rates existed in the vicinity 

of acid recovery equipment from the buildup of ruthenium and carryover from the waste evaporators. 

The ARC equipment was known to leak during operation. A spill in the ARC in 1967 required removal 

of nearly 24 in. of the affected floor. Six additional inches of concrete had to be poured over the 

original floor thickness to provide shielding. The spill also affected the cells on the ground level 

below, and reached the soil beneath the process building. It was determined to be the source of a 

groundwater plume that was discovered years later (WVNSC 1994 and WVDP 2005). 

A1.1 RADIOLOGICAL CONDITIONS

A1.1.1

In November 1967 piping from the Acid Recovery Cell read 3.5 R/hr at 12 in. 

External Exposure Data

On March 5, 1968, a tank in the Acid Recovery Cell read 30 R/hr on its sides and 100 R/hr at the 

bottom (Wenstrand 1968b). 

In April 1968 chemical decontamination of the Acid Recovery Cell reduced exposure rates from a 

general level of 2 R/hr to 4 R/hr to 125 mR/hr. The maximum exposure rate was reduced from 

50 R/hr to 3 R/hr. Shielding of localized activity reduced the general level further to 120 mR/hr (Keely 

1968j). 

Decontamination efforts in the Acid Recovery Cell in September 1969 reduced radiation levels by a 

factor of 20 (Wenstrand 1969a). 

A2.0 ACID RECOVERY PUMP ROOM

The ARPR was located in the southwest corner of the plant on the ground level. It contained 

equipment that supported operation of the Acid Recovery System and a sampler that collected a 24-

hour composite sample of the fractionator condensate. This condensate was discharged to the 

plant’s liquid effluent interceptor system. 

Decontamination work in the ARPR was performed between April 28, 1976, and May 6, 1976. On 

May 6, 1976, the door between the ARPR and the South Stairs was sealed using a plywood cover 

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with its edges caulked over. A concrete sill had been installed at the door on May 1, 1976 (WVNSC 

1994). 

A2.1 RADIOLOGICAL CONDITIONS 

The ARPR was affected by the high exposure rates associated with the acid recovery system 

throughout the plant and was chronically contaminated from maintenance operations. In addition, 

leaks from ARPR equipment resulted in erosion of the concrete walls and floor. 

A2.1.1

In early February 1967 operators taking samples in the ARPR had only 15 minutes before they 

reached their allowable weekly exposure limit. By February 19, 1967, exposure rates had increased 

to the point where the operators had only 2 minutes before they reached their weekly limit (Keely 

1967c). 

External Exposure Data

“Easily” removable contamination continued to be a problem in the ARPR (and the OGBR) in April 

1968, giving skin dose rates of 120 mrad/hr (Keely 1968j). 

In February 1972 replacement of the 7G-5 jet in the ARPR required 40 rem of whole-body exposure. 

The initial working area exposure rate was 60 R/hr. Shielding was installed to reduce the exposure 

rate to where a reasonable working time could be established (Wenstrand 1972f). 

An exposure rate of 30 R/hr was encountered in the ARPR in September 1972 (Morrow 1972a). 

In October 1972 exposure rates of 20 R/hr were encountered during cleanup work in the ARPR. This 

is referred to as an area radiation dose rate. It was not a hot spot (Morrow 1972b).

A2.1.2

Smears from the ARPR in February 1968 showed typical levels of 1.8E+06 dpm/100 cm2 beta 

(Wenstrand 1968e).

Removable Contamination Data

A3.0 ANALYTICAL AISLE, LABORATORIES, AND NORTH OPERATING AISLE

The analytical level of the plant (131-ft elevation) included six radiochemical labs:

• Hot Lab

• QC Lab

• Alpha Lab

• Product Lab (also sometimes referred to as the Uranium Lab)

• Mass Spectrometry Lab

• Emission Spectrometry Lab

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There was also a counting room north of the Mass Spectrometry Lab and a chemical storage room. 

Access to the laboratory areas was via the Analytical Aisle. The Analytical Aisle contained viewing 

windows and manipulators for operations within the analytical hot cells and Sample Cell 2C. 

Under the WVDP the laboratories were modified somewhat to support HLW vitrification operations. 

Some of the laboratories were expanded and renamed. One was called the Radiochemistry Lab. 

Another was the Vitrification Lab. 

The North Operating Aisle was on the north end of the analytical level east of the CPC Crane Room

(CCR). On the east side of the North Operating Aisle was the access to the hoist mechanism for the 

shield door between the PMC and the Process Mechanical Cell Crane Room (PMCR). This area was 

known as the penthouse. 

NFS decontaminated all of the operating era laboratories after shutdown. 

A3.1 RADIOLOGICAL CONDITIONS

A3.1.1

A chronic source of exposure in the Analytical Aisle was activity buildup in the ventilation system air 

washer. Activity in the washer caused exposure rates up to 40 mR/hr in the Analytical Aisle, which 

was located directly above the washer cell. NFS (1976) states that this exposure rate could be kept at 

20 mR/hr with repeated decontamination of the washer and the addition of lead shielding to the floor 

of the Analytical Aisle. 

External Exposure Data

In addition to the air washer, the ventilation system ducting was another chronic source of external 

personnel exposure in the analytical laboratories. Buildup of activity in the duct would produce 

exposure rates in the analytical laboratories of 10 mR/hr to 20 mR/hr. Lead shielding was added to 

the floors of the labs to the extent allowed by floor loading constraints (NFS 1976). 

Another source of exposure was buildup of contamination from maintenance activities in the 

penthouse area resulted in exposure rates of 5 mR/hr to 10 mR/hr in the adjacent analytical areas

(Wenstrand 1972a). 

Routine handling of samples was also an exposure mechanism. Samples having dose rates up to 

500 mrem/hr were analyzed in the laboratories in hoods. Shadow shielding was apparently used 

(NFS 1973b, p. 6.1-6).

In early 1968 the Hot Lab contained a high radiation area caused by solution leaking from a faulty 

drain (Runion 1968, p. 21 of Action Plan - Part 2). 

In October 1969 the exposure rate in the Analytical Aisle was 50 mR/hr. 

In 1990 dose rates in the North Operating Aisle ranged up to 8 mrem/hr. In 2004 the general level 

was around 1.5 mrem/hr, and in 2005 it was typically less than 1 mrem/hr except for one area by the 

east doorway that ranged up to 2 mrem/hr (WVDP 2005). 

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As of 2005 dose rates in the analytical area were low, coming primarily from fume hoods and the 

ventilation duct. Typical levels were said to be 0.1 mrem/hr to 2 mrem/hr, with a level of 2.9 mrem/hr 

being reported in the Analytical Aisle in May of 2005 (WVDP 2005). 

A3.1.2

On November 29, 1966, particulate activity from the PMC leaked into the analytical chemistry 

laboratories resulting in “gross contamination of all laboratories.” Cleanup took 4 days. 

Removable Contamination Data

An inspection by the plant housekeeping committee on January 27, 1967, pointed out a Zone IV leak 

in the Analytical Aisle that was asserted to have been there for months. 

In June 1969 a sample bottle exploded in the Alpha Lab causing removable contamination levels of 

up to 2E+07 dpm/100 cm2 alpha. The report states, “No personnel internal exposure to contamination 

occurred” (Keely 1969d). 

A4.0 CELL ACCESS AISLE

The Cell Access Aisle (CAA) was on the ground level north of Extraction Cells 2 and 3 and west of the 

UPC. It contained shield doors that provided access to Extraction Cells 2 and 3, the Liquid Waste 

Cell, and the UPC. The CAA served as an airlock when contact maintenance was being performed in 

these areas (WVDP 2005). 

A4.1 RADIOLOGICAL CONDITIONS

A4.1.1

Smears from the CAA taken January 15, 1968, showed typical removable levels of 

5,000 dpm/100 cm2 alpha (Wenstrand 1968f).

Removable Contamination Data

In June 1968 the typical removable contamination in the CAA was said to remain at 800,000 dpm/100 

cm2 alpha (Keely 1968a). The same level is reported for July 1968 also (Keely 1968k). 

In February 1969 smearable contamination in the CAA was reduced from 200,000 dpm/100 cm2 “to a 

low Zone 3” (Keely 1969f). 

A4.1.2

The CAA became an airborne contamination area during January 1967 (Keely 1967d, p. 2). 

Airborne Contamination Data

A5.0 CELL CRANE ROOMS

The remotely operated cranes and power manipulators used in the PMC, CPC, and GPC required 

periodic maintenance. Access to this equipment was via crane rooms located at one end of each cell. 

A shield door separated the crane rooms from the adjoining process cells. The equipment was first 

remotely decontaminated using spray systems before contact maintenance was performed in the 

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crane rooms. Rubber gloves and mats were sometimes used to reduce beta exposures in these 

areas (NFS 1976). 

A5.1 IMPACT OF THE THOREX CAMPAIGN ON AIRBORNE ACTIVITY LEVELS IN THE 

CELL CRANE ROOMS

In November 1968 airborne activity in the cell crane rooms was noted to be increasing due to alphaemitting thorium daughters from processing of thorium-uranium fuel from Indian Point 1. NFS 

developed methods to correct count data to determine the concentration of the long-lived alpha 

emitters (Keely 1968f). By December 1968 airborne activity in the cell crane rooms was said to 

remain “at maximum levels” (1,000 MPC) during Con-Ed (Indian Point 1) Core A processing “because 

of short-lived Thoron decay product.” Occasional water fogging was used to reduce airborne levels 

for entries (Keely 1969a). In February 1969 the airborne activity in the crane rooms “remains at 

maximum levels” (1,000 MPC) “as thorium contamination in cells still provides a source for short-lived 

alpha-emitting daughters.” The maximum concentration during the month was 50,000 MPC in the 

CCR (Keely 1969f). 

A5.2 GENERAL PURPOSE CELL CRANE ROOM AND EXTENSION

The GCR was directly west of the GPC on the same level (below grade). In 1971 the GCR was 

lengthened to accommodate storage of the crane and the power manipulator. This area is referred to 

as the GCR Extension. 

The GCR was used for contact maintenance work on the GPC crane, the GPC power manipulator, 

and the mechanism for the door between the cell and the GCR. This equipment would typically be 

decontaminated before maintenance activities (WVNSC 1994). Water would reportedly infiltrate the 

area on occasion during heavy rains (WVDP 2005).

In 2001 the GCR was used by the WVDP to support clean-up activities in the GPC (WVDP 2005). 

A5.2.1

A5.2.2

Radiological Conditions

During operations, the typical and maximum exposure rates observed in the GCR before

decontamination were 8 R/hr and 25 R/hr, respectively (NFS 1976). 

External Exposure Data

In 1999, the general area dose rate in the GCR was approximately 100 mrem/hr, except for the south 

side, which showed approximately 140 mrem/hr. The highest level measured was 2,200 mrem/hr on 

the floor near the sump (WVDP 2005). 

In 2002 the general dose rate in the GCR extension was in the range of 15 mrem/hr to 30 mrem/hr 

(WVDP 2005). 

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A5.2.3

A smear survey of the GCR airlock on February 24, 1968, showed removable beta levels ranging from 

11,000 dpm/100 cm2 to 200,000 dpm/100 cm2 (Wenstrand 1968e, p. 3).

Removable Contamination Data

Removable contamination surveys in the GCR in 1999 and 2000 showed levels up to 28,000 dpm/100 

cm2 alpha and 1.25E+6 dpm/100 cm2 beta (WVDP 2005). 

A5.2.4

In January 1968 airborne activity in the GCR exceeded 1000 MPC on 3 successive days (January 23

to 25). The maximum concentration during this time was 30,000 MPC, or 6E-08 μCi/cm3 alpha 

(Wenstrand 1968e, p. 2).

Airborne Contamination Data

On February 6, 1968, airborne activity in the GCR was 9.4E-11 μCi/cm3 alpha and 1.2E-8 μCi/cm3

beta (Wenstrand 1968e, p. 3).

In June 1968 airborne activity in the GCR remained above 1000 MPC for 3 days, preventing entry 

(Keely 1968a). (1000 MPC was NFS’ limit for supplied air respiratory protection.) 

A5.3 PROCESS MECHANICAL CELL CRANE ROOM

Access to the PMCR, which was on the 114.5-ft elevation of the plant, was via a hatch in the ceiling of 

the MRR below. 

A5.3.1

During operations, the typical and maximum exposure rates observed in the PMCR before

decontamination were 7 R/hr and 20 R/hr, respectively (NFS 1976). 

External Exposure Data

In 2003 the general area dose rate in the PMCR ranged from 45 mrem/hr to 100 mrem/hr 3 ft above 

the floor (WVDP 2005). 

A5.3.2

On February 13, 1968, removable beta contamination in the PMCR airlock ranged from 16,000 

dpm/100 cm2 to 600,000 dpm/100 cm2

 (Wenstrand 1968e, p. 3). 

Removable Contamination Data

In 2003 removable contamination on the PMCR door hoist mechanism was up to 14,000 dpm/100 cm2

alpha and 263,000 dpm/100 cm2 beta. No data were available for the Crane Room itself, but the 

levels were expected to be high (WVDP 2005). 

A5.3.3

In February 1968 “typical” airborne activity in the PMCR was 6.6E-7 μCi/cm3 beta and 8.2E-10 

μCi/cm3 alpha (Wenstrand 1968e, p. 2). 

Airborne Contamination Data

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A5.4 CHEMICAL PROCESS CELL CRANE ROOM

The CCR was at the north end of the CPC oriented 90 degrees to its west wall. It was two levels 

above the ground level of the CPC on the 131-ft elevation. 

A5.4.1

During operations, the typical and maximum exposure rates observed in the CCR before

decontamination were 3 R/hr and 6 R/hr, respectively (NFS 1976). 

External Exposure Data

A5.4.2

On February 13, 1968, typical removable beta contamination in the CCR airlock ranged from 120,000 

dpm/100 cm2 to 400,000 dpm/100 cm2

 (Wenstrand 1968e, p. 3). 

Removable Contamination Data

A5.4.3

On February 23, 1968, airborne activity in the CCR was 1.5E-10 μCi/cm3 alpha and 7.2E-7 μCi/cm3

beta (Wenstrand 1968e, p. 3).

Airborne Contamination Data

During February 1969 the concentration of thoron decay progeny in the CCR reached a maximum of 

50,000 MPC (Keely 1969f). This was despite processing of thorium fuel from Indian Point Unit 1 

having been completed on January 20, 1969. 

A6.0 CHEMICAL OPERATING AISLE AND CHEMICAL VIEWING AISLE

The COA was east of the CPC on the 114.5-ft elevation, between the CPC and the PMC. The 

Chemical Viewing Aisle (CVA) was west of the CPC, between it and the 2nd floor offices in the Office 

Building. 

The COA contained Sampling Station 1-C, which was used to remotely sample the process vessels in 

the CPC. 

A6.1 RADIOLOGICAL CONDITIONS

A6.1.1

The COA contained a high radiation area caused by backup of solution in a steam line (Runion 1968, 

p. 16 of Action Plan - Part 2). 

External Exposure Data

A July 27, 1972, summary of plant decontamination activities said radiation levels in the COA had 

been reduced through various decontamination efforts, both in the COA and in the extraction cells. 

Remaining sources of radiation were contamination on the floor and residual activity in piping and 

instrument lines. A steam line was said still to require heavy shielding and should be replaced to 

reduce radiation levels further. The report also mentions exposure rates in the CVA originating from 

the Southwest Stairwell have not changed from a previous survey in March 1972 (Wenstrand 1972d, 

p. 2). 

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A6.1.2

Contamination would be carried into Sampling Station 1-C from the manipulator used to perform the 

remote sampling. This contamination was said to be limited to the immediate vicinity of the sampling 

station (Runion 1968, p. 16 of Action Plan - Part 2). 

Removable Contamination Data

A7.0 CHEMICAL PROCESS CELL 

The CPC is where the leaching (dissolving) process was performed on small pieces of fuel following 

mechanical segmentation in the PMC. The CPC contained two 4,000-L dissolvers. The dissolved 

fuel was chemically adjusted as necessary in the CPC to create the feed solution for the solvent 

extraction process carried out in the extraction cells.

A7.1 RADIOLOGICAL CONDITIONS

The CPC was never a routinely occupied area. 

In May 1968 a remote survey just inside the CPC showed 220 R/hr. (Keely 1968e). In 1981 the 

exposure rate in the CPC was said to range from 12 R/hr to 32 R/hr (Golden, Frank, and Prina 1982).

A8.0 CONTROL ROOM

The Control Room was located on the fourth floor of the processing plant on the 144-ft elevation. It 

was used for the monitoring and control of plant processes. Adjacent to the Control Room were the 

Control Room Office and an area containing water distillation equipment. 

A8.1 RADIOLOGICAL CONDITIONS

There were two chronic sources of external personnel exposure in the Control Room:

• The plant’s main ventilation duct

• The recovered acid storage tanks in the Hot Acid Cell

A8.1.1

In January 1967 exposure rates in the Control Room were 10 mR/hr to 15 mR/hr due to activity in the 

ventilation duct (between the air washer and the stack) (Keely 1967d, p. 3). 

External Exposure Data

On February 26, 1968, the exposure rate in the Control Room ranged from 1 mR/hr to 5 mR/hr 

(Wenstrand 1968e, p. 2).

Exposure rates in the Control Room increased from 2 mR/hr to 10 mR/hr in June 1969 (Keely 1969d). 

The exposure rate dropped back to 2.5 mR/hr in July after flushing of the ventilation duct (Keely 

1969e).

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The July 27, 1972, summary of plant decontamination activities says reduced activity in the ventilation 

duct (post shutdown) resulted in “some reduction” of radiation levels in the Control Room, however, a 

spill in the northwest corner was said to have increased levels in that area (Wenstrand 1972d, p. 3). 

NFS (1976) says the exposure rate in the Control Room from radioactivity in the ventilation ducting 

was typically 5 mR/hr to 10 mR/hr. 

In May of 2005 dose rates in the Control Room area were generally less than 1 mrem/hr with a high of 

1.4 mrem/hr (WVDP 2005). 

A9.0 EQUIPMENT DECONTAMINATION ROOM AND EQUIPMENT DECONTAMINATION

ROOM VIEWING AISLE

The EDR was on the ground level in the northwest corner of the process building, adjacent to the 

CPC and the Scrap Removal (SR) area. It contained a decontamination pool known as the “soaking 

pit” and equipment for decontamination of large vessels and equipment. 

A9.1 RADIOLOGICAL CONDITIONS

A9.1.1

The general area exposure rate in the EDR in the first quarter of 1968 was 30 mR/hr to 40 mR/hr. 

Extremity dose rates were estimated at 590 mrad/hr to 650 mrad/hr (Urbon 1968b). 

External Exposure Data

A9.1.2

Removable beta contamination in the EDR ranged up to 2.1E+06 dpm/100 cm2 during the week of 

February 12, 1968 (Wenstrand 1968e, p. 3).

Removable Contamination Data

In September 1968 the EDR was reduced to “Zone 3” levels for removable contamination during the 

month (meaning it was Zone IV previously) (Keely 1968c). 

A9.1.3

Airborne activity in the EDR exceeded 1,000 MPC on one sample during March 1968 (Wenstrand 

1968b). 

Airborne Contamination Data

A10.0 EXTRACTION AND PRODUCT PURIFICATION CELLS 

Solvent extraction was carried out in a series of four cells: Extraction Cells 1 to 3 (XC1 to XC3), and 

the Product Purification Cell (PPC). The four cells were approximately 60 ft high each, so they 

passed through every elevation of the process building from ground level up to the 160-ft elevation. 

The cells contained extraction columns and evaporators. 

The first two partition cycles, separation of uranium and plutonium from fission products, followed by 

separation of the plutonium from the uranium, were carried out in XC1. The final partition cycle and 

first uranium and plutonium extraction cycles were conducted in XC2. The second (final) plutonium 

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and uranium extraction cycles and cleanup were performed in XC3. Final purification of plutonium 

and uranium nitrate process streams was performed in the PPC. Uranium was processed on the 

north side of the PPC and plutonium on the south side. 

Sometimes the plant records will contain references to the Plutonium Purification Cell or just the 

Plutonium Cell. These are references to the south (plutonium) side of the PPC. 

A10.1 RADIOLOGICAL CONDITIONS

The extraction cells presented both internal and external radiation exposure hazards from high 

airborne contamination levels and high dose rates. The PPC was subject to “significant” spills of 

radioactive liquids including one that contaminated the south wall of the south (plutonium) side to a 

height of 35 ft (WVDP 2005). During the operations era, entries into the PPC would cause airflow 

reversal issues on other levels of the plant, resulting in area contamination. 

A10.1.1

Surveys in XC3 on February 29, 1968, showed working exposure rates from 300 mR/hr to 500 mR/hr, 

with a localized reading of 5 R/hr at 12 in. from a particular component (Wenstrand 1968e, p. 2). 

External Exposure Data

In October 1969 a survey at the entrance to XC2 showed a whole-body exposure rate of 5 R/hr. 

A manned entry was made into XC1 on June 29, 1972, following water flush activities. A maximum 

exposure rate of 50 R/hr was indicated near tank 4D-1 (WVNSC 1994). Several more entries were 

made into XC1 in July 1972, identifying exposure rates up to 20 R/hr at Tank 4-D-1 (Morrow 1972c). 

A survey in XC1 on February 21, 1973, found exposure rates 30 R/hr to 50 R/hr at the second stair 

landing and 10 R/hr to 50 R/hr at the first landing (WVNSC 1994). 

Decontamination efforts in XC2 in May 1972 reduced exposure rates from the floor from 100 R/hr to 3 

R/hr (Wenstrand 1972g). 

The contamination on the south wall of the south side of the PPC read up to 53 mrem/hr as of 2003. 

The dose rate on the north side ranges from 1 mrem/hr to 3 mrem/hr (WVDP 2005). 

A10.1.2

On February 29, 1968, removable alpha contamination in XC3 ranged from 1.5E+06 dpm/100 cm2 to 

6.5E+06 dpm/100 cm2

. Removable beta ranged from 3.5E+06 dpm/100 cm2 to 6.4E+06 dpm/100 cm2

(Wenstrand 1968e, p. 3). 

Removable Contamination Data

A survey in 1986 showed a maximum of 400,000 dpm/100 cm2 removable alpha on the floor of the 

north side of the PPC before application of strippable coating. Levels on the walls of the north side 

showed a maximum of 6,700 dpm/100 cm2 removable alpha. Removable levels in the PPC remained 

high as of 2003. 

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A10.1.3

Air samples from the PPC showed concentrations of 2.0E-9 μCi/cm3 alpha on February 23, 1968, and 

6.7E-11 μCi/cm3 alpha on February 26, 1968 (Wenstrand 1968e, p. 2). 

Airborne Contamination Data

In April 1968 modification work in the PPC (plutonium side) had to be interrupted on several 

occasions due to excessive airborne activity levels (>1,000 MPC). Smearable contamination reached 

5E+06 dpm/100 cm2 alpha (Keely 1968j). On April 19, 1968, the concentration in the PPC reached 

5000 MPC (Author unknown 1968). 

Airborne activity in the PPC reached 30,000 MPC during an entry on October 30, 1968. Personnel 

had to be evacuated until remote decontamination reduced the level to 1000 MPC (Keely 1968d). 

An entry was made into the PPC in May 1969 after an isotopic evaluation showed the plutonium 

fraction of the airborne contamination was 4% before cell entry, 30% after resin dump, and 7% at 

completion. Entry had been prevented at first when an assumption of 100% plutonium for the alpha in 

air meant the concentration exceeded limits for supplied air respiratory protection (Keely 1969c, p. 1). 

A November 11, 1971, memorandum alludes to the problems of airborne plutonium levels in the PPC 

and that workers have to wait for it to settle before making entries. Surface contamination and 

resuspension are also issues (Heacker 1971). 

A11.0 EXTRACTION CHEMICAL ROOM

The XCR was above the extraction cells on the 160-ft elevation of the plant. It is where chemical 

solutions for the solvent extraction process were prepared. 

A11.1 RADIOLOGICAL CONDITIONS

Some steps of the extraction chemical makeup process made use of acid from the acid recovery 

system. High exposure rates were therefore an issue. In addition, the XCR was chronically 

contaminated from leaks, spills, and equipment maintenance or replacement activities. 

A11.1.1

The typical exposure rate in the XCR during the month of February 1968 ranged from 2.5 mR/hr to 

500 mR/hr (Wenstrand 1968e, p. 2).

External Exposure Data

Dose rates of 50 R/hr were measured in the vicinity of the XCR in December 1971. The source was 

recovered acid in a tank (Wenstrand 1972h). 

The July 27, 1972, summary of plant decontamination activities says flushing of tanks and removal of 

contaminated concrete in the XCR had reduced radiation levels by a factor of 10 to 100. The principal 

remaining radiation source was a tank reading 20 mR/hr to 30 mR/hr at contact. This tank read 1 R/hr 

to 2 R/hr before decontamination (Wenstrand 1972d, p. 3). 

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A11.1.2

Removable beta under the tanks in the XCR ranged from 21,000 dpm/100 cm2 to 240,000 dpm/100 

cm2 on January 22, 1968 (Wenstrand 1968f, p. 3). 

Removable Contamination Data

Smearable contamination in the XCR area was well into Zone 4 levels in December 1971 (Wenstrand 

1972h).

A12.0 EXTRACTION S AMPLE AISLE AND PROCESS SAMPLE CELL 2

The XSA was located on the 131-ft elevation of the plant, which was known as the analytical level. 

The XSA was adjacent to the north walls of the four extraction cells, between the cells and the 

laboratories. It included PSC-2, which was equipped with gloveboxes and sampling stations for 

sampling the adjacent extraction cells. The gloveboxes were removed in 1986 (WVDP 2005). 

A12.1 RADIOLOGICAL CONDITIONS

The XSA contained two sample stations that suffered from contamination problems purported to be 

from insufficient bagging and airflow reversals caused by entries into the PPC (Runion 1968, p. 14 of 

Action Plan - Part 2). The contamination problems from entries into the PPC were such a problem 

that, as of early 1968, NFS required the XSA to be cleared of personnel and the door locked before a 

PPC entry was made. Subsequently, the XSA had to be surveyed once work in the PPC was 

completed (Runion 1968, p. 15 of Action Plan - Part 2). At the time they were enacted (early 1968), 

these measures were intended to be temporary until the sampling stations and gloveboxes could be 

improved to mitigate the contamination problems. 

A12.1.1

General removable alpha levels in the XSA ranged from 5,000 dpm/100 cm2 to 10,000 dpm/100 cm2

on November 20, 1967. 

Removable Contamination Data

A12.1.2

In October 1967 air samples “taken on each shift” from the XSA and PSC-1 showed long-lived 

airborne alpha concentrations of 2E-12 μCi/cm3 and 5.4E-11 μCi/cm3

 (Wenstrand 1967d). 

Airborne Contamination Data

Airborne activity levels rose to MPC in the XSA during a shutdown of the ventilation system following 

the filter failure incident on September 4, 1968 (Keely 1968c).

A13.0 FUEL RECEIVING AND STORAGE

The FRS facility was used to handle and decontaminate shipping casks for nuclear fuel and to move 

fuel using service crane bridges. It consisted of the fuel SP, the cask receiving area, the cask 

decontamination area (pit), and the pool water treatment area. Fuel was transferred from the FRS to 

the PMC via a transfer tunnel in the southwest corner of the SP. 

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Other facilities associated with the FRS were a radioactive waste processing building (known as the 

Hittman Building) and the resin pit.

The radioactive waste processing building was located in a yard area north of the FRS. It was used 

for dewatering and storage of filter media from FRS pool cleanup systems and ion exchange resins. 

Spent resin was dewatered and stored in High Integrity Containers located 50 ft north of the FRS 

Building (WVDP 2005). 

The resin pit was an in-ground concrete vault located outside the southeast corner of the FRS 

Building. Spent filter media and ion exchange resin from the pool water treatment area were flushed 

to a drum in the pit. A drain line went from the pit to the Interceptor system. 

Fuel was stored in the FRS pool until 2001. In 2002 and 2003 the pool was drained, empty canisters 

and storage racks were removed, and the SP and Cask Unloading Pool (CUP) were decontaminated 

and sealed (WVDP 2005). 

A13.1.1

Receipt and storage of ruptured fuel assemblies resulted in significant buildup of fission products in 

the SP. The radioactivity concentration in the pool eventually reached 0.1 μCi/cm3

, prompting the 

installation of auxiliary cleanup systems to augment the primary system. However, even with the 

auxiliary system the contamination level of the pool water could only be reduced to 0.01 μCi/cm3

. 

Hence, significant levels of contamination would be found “Whenever items were removed from the 

pool.” Radiation levels would reach 20 mR/hr to 100 mR/hr during routine operations in some areas 

of the FRS (NFS 1976).

Radiological Conditions

Personnel exposures in the FRS area would also result from preparation of casks for loading into the 

pool. Workers were exposed to the surface dose rates from the casks while they were prepared for 

immersion (NFS 1976). Incoming casks could have a contact dose rate up to 200 mR/hr (limit per 

shipping regulations). Personnel were routinely in close proximity to the casks while preparing them 

for immersion into the pool and unloading of the fuel (Runion 1968, p. 1 of Action Plan - Part 2). FRS 

workers removing and cleaning fuel shipping cask internals worked in exposure rates up to 300 mR/hr 

(Loud 1965b, p. 2). The average collective dose for FRS workers over 1970 and 1971 was 250

person-rem. A May 12, 1975, Nuclear Insurance Survey report by Johnson & Higgins states the 

highest level of personnel exposure at West Valley was around 400 mrem/quarter received by the fuel 

unloading operators (Johnson & Higgins 1975). 

Liquids would be spilled into work areas in the FRS, resulting in airborne contamination and personnel 

intakes. There was also potential for exposure to fission gases released when cask lids were 

removed (in the pool) (Bailey 1990). Other routes of personnel exposure in the FRS were handling of 

contaminated tools, drumming of filter and demineralizer media from the pool water treatment system, 

and shipping cask decontamination operations. There were multiple instances of personnel 

exposures associated with flushing of shipping casks. 

Underwater tools and the crane cables and blocks allowed highly contaminated pool water to drip 

onto the service bridges. Casks were contaminated to “several hundred thousand” dpm/100 cm2 beta 

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from the pool water upon being removed from the pool. Unloaded casks were moved from the SP to 

the decontamination pit, where they were cleaned using spray and scrub techniques to return them to 

acceptable contamination levels for shipping. The FRS service crane bridges were repeatedly 

contaminated from FRS operations. The bridges were isolated and controlled via rope and step-off 

pads (Runion 1968, p. 1 of Action Plan - Part 2). 

A13.1.2

Table A-1 shows gross beta and gross alpha concentration data for the FRS CUP and SP for 

December 1965. 

Pool Water Concentration Data

Table A-1. Gross beta and gross alpha 

concentration data for the CUP and SP, 

December 1965. 

Pool

Concentration (μCi/cm3)

Gross beta Gross alpha

CUP 3.1E-5 2.5E-9

SP 2.8E-5 4.8E-9

Table A-2 shows concentration data for the CUP and SP for January to June 1966. Data are given for 

soluble gross beta, gross alpha and tritium; and insoluble gross beta and gross alpha. 

Table A-2. Radioactivity concentration data for the CUP and SP, January to June 1966. 

Date Pool

Soluble (μCi/cm3) Insoluble (μCi/cm3)

Gross beta Gross alpha Tritium Gross beta Gross alpha

January 1966 CUP 8.0E-5 1.6E-7 6.5E-7 8.8E-7 1.0E-8

SP 8.8E-5 3.0E-8 5.0E-6 2.0E-6 1.0E-8

February 1966 CUP 2.7E-5 < 1E-9 6.0E-6 1.9E-6 2.6E-9

SP 2.8E-5 3.7E-9 9.5E-6 1.4E-6 9.4E-9

March 1966 CUP 5.9E-5 3.3E-7 3.9E-6 5.5E-6 < 1E-8

SP 6.8E-5 < 1E-8 4.3E-6 2.0E-6 < 1E-8

April 1966 CUP 4E-5 1E-8 1E-5 3E-6 7E-10

SP 5E-5 < 1E-8 9E-6 3E-6 4E-9

May 1966 CUP 2E-4 6E-8 1E-4 6E-6 9E-9

SP 2E-4 5E-8 1E-4 5E-6 1E-9

June 1966 CUP 1E-4 < 1E-8 4E-4 4E-5 9E-9

SP 2E-4 < 1E-8 3E-3 1E-5 6E-10

For the February 1966 samples, “Gamma analysis showed that Co-60, Co-58, and Cs-137 are the 

principal contaminants” (Loud 1966b, p. 2). 

Table A-3 shows gross beta concentration data for the CUP and SP over the period September 1965 

to October 1969. Values reported for a month (rather than a specific date) are typical or 

representative values. Values are for the SP unless otherwise specified. 

After reprocessing operations ceased in 1972, the FRS pool was drained and cleaned. After refilling, 

the contamination level of the pool water was reduced to 5E-3 μCi/cm3 (NFS 1976).

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Table A-3. Gross beta concentration data for the CUP and SP, September 1965 to October 1969. 

Date

Gross beta

concentration

(μCi/cc) Remarks

September 1965 1.1E-04 CUP 

September 1965 7.8E-05 SP 

September 16, 1965 1.6E-04 CUP concentration due to ruptured fuel assembly. Previously was 5.4E-6 μCi/cc. 

October 1965 3.1E-04 CUP 

October 1965 2.8E-04 SP 

November 1965 1.4E-04 "No detectable alpha" 

August 1966 2.0E-03 Small cracks noted in pool, causing contamination on nearby floor areas. 

July 1, 1967 4.0E-03 Maximum. Low was 9E-5 μCi/cc 

September 6, 1967 7.5E-03

October 8, 1967 1.8E-02

March 1969 1.0E-01 Waiting for removal of ruptured N Reactor fuel before attempting cleanup. 

April 1969 1.0E-01

May 1969 1.2E-01

June 1969 1.2E-01

July 1969 8.0E-02

August 1969 1.7E-02

September 1969 8.0E-03

October 1969 4.0E-03

On May 11, 1972, isotopic measurements were made on FRS pool water to examine the 

effectiveness of cleanup systems. Concentration data were given for 134Cs, 137Cs, 106Ru/Rh, 144Ce/Pr, 

and 125Sb (Jaroszeski 1972). 

A13.1.3

Table A-4 shows exposure rates over the top of the FRS pools for November 1965 to May 1969.

External Exposure Data

Table A-4. Exposure rates over the top of the FRS pools, November 1965 to 

May 1969. 

Date

Exposure rate 

(mR/hr) Remarks

November 1965 0.2 

January 1966 0.1 to 0.15 

February 1966 0.1 to 0.15 Seven fuel casks were received during the month.

March 1966 0.1 to 0.15 Ten fuel casks were received during the month.

April 1966 0.2 Fifteen fuel casks were received during the month.

May 1966 0.2 

June 1966 0.4 

May 1969 30 whole body 

100 skin 

Average working level on bridges above pool.

In December 1965 the FRS pool demineralizer was placed in service “for a few days” and showed an 

exposure rate “at the side” of 1.5 R/hr (Loud 1966a). 

In September 1969 an expansion box on a fuel cask received read 5 R/hr on contact. Investigation 

found it contained high-activity primary coolant. Attempts to clean up the coolant using an ion 

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exchange unit were hampered when the unit read 350 R/hr after 3 hours of operation (Wenstrand 

1969a). 

On December 5, 1969, a line venting the cask leaked coolant onto the concrete outside the CUP, 

which resulted in a high radiation area that persisted for days. The contaminated concrete had to be 

removed and lead fill added (Duckworth 1970). 

In November 1971 shrouds were removed from the FRS pool resulting in exposure rates up to 40 

R/hr. Shielding was provided for the FRS crane operator, reducing his work area exposure rate to 

300 mR/hr (Wenstrand 1971c, p. 1). 

In June 1972 decontamination efforts on the CUP reduced exposure rates to approximately 5 R/hr 

(Wenstrand 1972i). 

During October 1972 exposure rates of 20 R/hr were encountered while handling spent resin in the 

FRS. This is referred to as an “area radiation dose rate”; that is, it was not a hot spot (Morrow 1972b).

During November 1972 exposure rates of 30 R/hr were encountered during removal of wastes from 

the FRS (Morrow 1972d). 

Representative exposure rates for the FRS in 2002 and 2003 were (WVDP 2005). 

• SP walls and floor = 6.7 mrem/hr to 8.5 mrem/hr.

• CUP walls and floor = 25.7 mrem/hr to 500 mrem/hr.

• Water treatment area = 5.4 mrem/hr on face of shield wall; 195 mrem/hr on middle of 

demineralizer column.

• Cask Decontamination Area = 0.17 mrem/hr in stall; 24 mrem/hr in pit sump.

• Hittman Building = 0.5 mrem/hr to 0.8 mrem/hr.

• Top of resin pit = 18 mrem/hr.

• Bottom of FRS pool transfer tunnel = 2,000 mrem/hr.

A13.1.4

Smears around the FRS Decontamination Pit showed removable levels from 3,000 dpm/100 cm2 to 

8E+06 dpm/100 cm2 beta on 1/24/1968 (Wenstrand 1968f, p. 3). 

Removable Contamination Data

Smearable contamination on an incoming cask was 1E+06 dpm/100 cm2 beta in July 1968 (Keely 

1968k). 

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In September 1969 smearable contamination in excess of 50,000 dpm/100 cm2 beta was found in the 

FRS rail track area (Wenstrand 1969a).

Smearable contamination on an incoming fuel cask showed a maximum of 7.5E+07 dpm/100 cm2

April 1971. This is assumed to be removable beta contamination, though the reference does not 

explicitly say so (Wenstrand 1971f, p. 2). 

NFS received fuel casks with removable contamination levels in excess of the 10 CFR Part 20 limit of 

22,000 dpm/100 cm2 five times between July 2 and November 7, 1974 (O’Reilly 1975, p. 1 of 

inspection report). 

After the decontamination activities in 2002 to 2003, removable contamination levels in the FRS area 

were less than 2000 dpm/100 cm2 alpha and 20,000 dpm/100 cm2 beta (WVDP 2005). 

A13.1.5

There was high airborne beta activity in the FRS on December 20, 1966, attributed to removable 

activity in a drain (Keely 1967e). 

Airborne Contamination Data

April 1969: The “maximum weekly” long-lived air activity above the FRS pool for April 1969 was 4E-10 

μCi/cm3 beta (Keely 1969g). 

January 1972: airborne concentrations in the FRS have decreased by a factor of 10 due to the 

cleanup effort (Wenstrand 1972b). 

A lid blew off a 55-gal drum during decanting of the FRS settling tank on May 15, 1974. Resulting 

airborne contamination levels were 1.56E-11 μCi/cm3 alpha and 2.66E-09 μCi/cm3 beta (G. 1974). 

A14.0 GUARD HOUSE

The Guard House was the main point of entry to and exit from the plant, and where employee film 

badges were stored. It was also referred to as the Gate House. 

A14.1 RADIOLOGICAL CONDITIONS

In August 1970 the AEC expressed concerned over exposure rates at the gatehouse. NFS said that 

while the exposure rates in that area were “higher than desired”, they did not consider it to be an 

unrestricted area vis-à-vis 10 CFR Part 20. The only persons routinely in that area were the guards, 

whom were described as being “low in radiation exposure.” NFS said radiation levels in that area 

would be “reduced to more desirable levels as soon as possible” (Duckworth 1970). The discussion 

pertains to the fact exposure rates at the Guard House exceeded 10 CFR Part 20 limits for an 

unrestricted area. 

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A15.0 GENERAL OPERATING AISLE

The GOA was below grade on the north side of the GPC. It contains four shielded viewing windows 

that were used by workers operating manipulators and other equipment in the GPC. 

A15.1 RADIOLOGICAL CONDITIONS

A15.1.1

A July 27, 1972, summary of plant decontamination activities says radiation levels in the GOA were 

reduced by a factor of two due to lower radiation levels in the HEV duct. It also says contamination 

around the west window was still causing a general area exposure rate of 20 mR/hr (Wenstrand 

1972d). 

Exposure Rate Data

The reported dose rate in the GOA as of 2005 was less than 1 mrem/hr (WVDP 2005). 

A15.1.2

The GOA was one of the areas that showed airborne contamination levels in excess of 40 MPC-hr

during flushing of the plant ventilation system. 

Airborne Contamination Data

A16.0 GENERAL PURPOSE CELL

The GPC was 25 ft below the ground level of the plant, and ran perpendicular between the PMC and 

the CPC. The GPC served to transfer baskets containing 200 kg (design basis) of segmented fuel 

between the PMC and the CPC for dissolving. Afterward, they would be moved back into the GPC so 

the remaining cladding and fuel hardware could be packaged for transfer into the SR Area for 

disposal. The GPC would also be used for packaging of other process wastes and damaged 

equipment for transfer into the SR Area.

The GPC had three shielded viewing windows in its north wall, denoted east to west as GPC-A, -B, 

and -C (Vance 1986). The west window was a source of contamination into the GOA, which ran 

adjacent to the north wall of the GPC. 

There was another cell on the other side of east wall of the GPC called the Miniature Cell. This cell 

was never used. 

Vance (1986) says discussions with former employees indicated there had been a metal fire in the 

GPC during the operations era when a batch of leached cladding spontaneously ignited when it was 

dumped. It is not known specifically when this occurred. WVDP (2005) says three cladding fires 

occurred during operations.

A16.1 RADIOLOGICAL CONDITIONS

The GPC was never a routinely occupied area. 

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A16.1.1

In 1986 dose rates 6 ft above the floor ranged from 40 rem/hr to 340 rem/hr. At 2 ft above the floor 

the dose rates ranged up to 650 rem/hr. 

External Exposure Data

A characterization survey performed in November 2004 showed the following dose rates:

• 3 ft above the floor = 4 rem/hr to 200 rem/hr.

• 6 ft above the floor = 3 rem/hr to 45 rem/hr.

• 9 ft above the floor = 2 rem/hr to 32 rem/hr.

A16.1.2

Smear samples collected in 1986 showed removable 137Cs up to 388 μCi/smear. 

Removable Contamination Data

A17.0 HEAD END VENTILATION BUILDING AND CRANE ROOM

The HEV Building was located on the east side of the MSM Shop. It was built in 1970 to 

accommodate the HEV system. The HEV Building had two levels. The lower level housed filters, 

blowers, ductwork, and other equipment. The upper level contained a crane and filter change 

equipment. This area was often referred to as the HEV Crane Room. Like the main ventilation 

system, the HEV system could be operated on steam or electric power. 

A17.1 RADIOLOGICAL CONDITIONS

A17.1.1

On October 12, 1971, extensive contamination was found in a surface water drainage ditch from the 

HEV area. Typical exposure rates ranged from 60 mR/hr to 100 mR/hr, and up to 150 mR/hr inside 

drains (Wenstrand 1971g). 

External Exposure Data

In July 1972 decontamination of the floor and dock area of the HEV Crane Room reduced radiation 

levels in these areas by a factor of 10 to 12. Radiation fields from “shielding flaws at the sides of the 

HEV” still required shielding, though levels were lower due to less activity in the filters and 

decontamination of the filter chambers (Wenstrand 1972d, p. 2). 

In 2002 dose rates from the HEV prefilters ranged up to 50 rem/hr. The roughing filters ranged up to 

18 rem/hr and the first stage high-efficiency particulate air (HEPA) filter bank up to 2.3 rem/hr. 

Ductwork in the adjacent operating aisle ranged from 3 mrem/hr to 50 mrem/hr. Dose rates up to 61 

mrem/hr were measured in the lower level of the HEV Building in May of 2005 (WVDP 2005). 

A17.1.2

No specific information has been identified. The 2005 draft of the WVDP Decommissioning Plan says 

removable levels in the HEV Building “are expected to be significant” (WVDP 2005). 

Removable Contamination Data

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A18.0 HOT ACID CELL

The Hot Acid Cell was located on top of the CPC at an elevation of 148 ft. It contained two storage 

tanks (7D-11 and 7D-12) for recovered acid. It also contained a pump in a niche. 

A18.1 RADIOLOGICAL CONDITIONS

A18.1.1

On February 14, 1994, exposure rates of 5 mR/hr to 10 mR/hr were present near the floor and 10 

mR/hr to 15 mR/hr at the base of the acid tanks (WVNSC 1994). 

External Exposure Data

In 2001 dose rates inside the Hot Acid Cell ranged from 0.3 mrem/hr to 80 mrem/hr. Dose rates from 

piping, valves and the floor drain ranged from 5 mrem/hr to 60 mrem/hr (WVDP 2005).

A18.1.2

A 1994 removable contamination survey showed levels up to 450 dpm/100 cm2 alpha and 42,000 

dpm/100 cm2 beta (WVDP 2005). 

Removable Survey Data

A19.0 HOT SHOP

The Hot Shop was a portion of the “old maintenance shop adjacent to the new laundry” (Runion 1968, 

p. 7 of Action Plan - Part 2). The area was used for decontamination, repair, and maintenance of 

equipment such as MSMs and power manipulators. The Hot Shop was provided in 1967 to permit 

repair of contaminated equipment. A new Decontamination and MSM Repair Facility was under 

construction as of early 1968 with completion expected July 1 (Clark 1968b). 

A19.1 RADIOLOGICAL CONDITIONS

Equipment was bagged before transfer to the Hot Shop, but maintenance activities there resulted in 

contamination of it and surrounding areas. Hot Shop personnel wore respiratory protection.

A19.1.1

On November 3, 1967, the Hot Shop showed a general background of 10 mR/hr to 20 mR/hr. Some 

manipulator parts read up to 21 rad/hr. 

External Exposure Data

A19.1.2

Smears of the Hot Shop on January 22, 1968, showed removable beta levels of 240,000 dpm/100 

cm2 to 4E+06 dpm/100 cm2 and alpha levels of 10,000 dpm/100 cm2 to 12,000 dpm/100 cm2

(Wenstrand 1968f, p. 3).

Removable Contamination Data

Typical removable beta contamination in the Hot Shop was 100,000 dpm/100 cm2 on February 27, 

1968 (Wenstrand 1968e, p. 3).

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Removable contamination increased above allowable limits in the Hot Shop on 3/3/1968 (Wenstrand 

1968b, p. 2).

A20.0 INSTRUMENT SHOP/OLD INSTRUMENT SHOP

The Instrument Shop was located in the Ventilation Supply Room, which was east of the PMC on the 

114.5-ft elevation. It was used for the repair of uncontaminated instruments until this function was 

relocated sometime in 1967. After this time the area was referred to as the Old Instrument Shop. The 

area itself was physically removed in 1985. 

A20.1 RADIOLOGICAL CONDITIONS

A20.1.1

In January 1967 exposure rates in the Instrument Shop were 10 mR/hr to 40 mR/hr from activity in the 

ventilation duct (between the air washer and the stack) which passed adjacent to the area (Keely 

1967d, p. 3). 

External Exposure Data

A survey in the Old Instrument Shop on February 26, 1968, showed exposure rates from 250 mR/hr to 

400 mR/hr (Wenstrand 1968e, p. 2).

As of 2005 dose rates adjacent to the ventilation duct in the Ventilation Supply Room ranged up to 7 

mrem/hr (WVDP 2005). 

A20.1.2

Removable contamination in the Old Instrument Shop was reduced from “high Zone 4 levels to low 

Zone 3 levels” during the month of April 1969 (Keely 1969g). However, a leaking duct recontaminated the area back to Zone 4 levels the following month (Keely 1969c, p. 1). 

Removable Contamination Levels

A21.0 INTERCEPTOR AND NEW INTERCEPTOR

The original Interceptor was located within the plant security fence just east of the process building. 

Its purpose was to collect liquid wastes from floor drains, sink drains, blowdown from process water 

and steam, and cask wash water from the FRS. These liquids were collected in batches and 

analyzed before transfer to the lagoon system. Wastes were then stored in Lagoons 1 and 2 until 

they were processed in the Low-Level Waste Treatment (LLWT) Facility. Before May 1971, the 

wastewater went from Lagoon 2 to Lagoon 3 before being discharged to surface streams. 

Potentially radioactive condensates in the processing plant were collected and sampled in batches 

before disposal. If the gross beta concentration exceeded 3E-07 μCi/cm3

, the waste was transferred 

to the Interceptor system. Otherwise, it was discharged directly to surface streams. Secondary 

cooling water used in the High Level Waste Storage Facility and in the FRS cooler was also 

discharged directly to surface streams. 

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A new Interceptor was installed by the WVDP in the late 1990s consisting of twin, open-top concrete 

storage tanks. The new Interceptor is located just east of the old one. The tanks are used to store 

radioactive liquids before treatment in the LLWT Facility (WVDP 2005). 

In 2005 the old Interceptor was being used for storage of radioactive liquids that exceeded the effluent 

limit of 5E-03 μCi/cm3

. Liquids can be transferred from the old Interceptor to the new Interceptor via 

steam jet (WVDP 2005).

A21.1 RADIOLOGICAL CONDITIONS

A21.1.1

Exposure rate surveys performed in August 2003 on the old Interceptor showed a range of 0.4 mR/hr 

at the top of the tank to 408 mR/hr near the bottom at a distance of 1 ft from the tank wall (WVDP 

2005). 

External Exposure Data

A valve in the new Interceptor valve pit showed 0.3 mR/hr in October 2003 (WVDP 2005). 

A22.0 LAUNDRY

A new laundry facility went into operation in 1967 in an effort to lower workplace contamination 

experienced by laundry workers.

A22.1 RADIOLOGICAL CONDITIONS

A22.1.1

As of early 1968 clothing articles with dose rates up to 50 mR/hr could be sent to the plant laundry. 

External Exposure Data

An AEC inspection report from 1972 summarizes external dose received by a laundry worker at West 

Valley between October 12, 1966, and January 19, 1972. His total whole-body dose over this period 

was 8.950 rem, and he had no quarterly dose in excess of 1.2 rem. The report states the worker’s 

“extremity and skin exposure” over the 5.25 years was 11.7 rad. It is not known why the two were 

summed, if that was the case (Nelson 1973a, p. 14 of attachment). The whole-body dose averages to 

426 mrem/quarter, or 1.7 rem/yr. 

A22.1.2

Waste from the laundry facility backed up in June 1967, contaminating the First Aid Room to Zone IV 

levels (Keely 1967f). 

Removable Contamination Data

A22.1.3

On February 9, 1967, the airborne alpha activity in the laundry area increased “to levels requiring 

personnel to wear masks” (Keely 1967a, p. 1). 

Airborne Contamination Data

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As of early 1968 laundry personnel wore canister masks and worked under a hood flowing at “about 

1,300 scfm” while sorting contaminated clothing before washing (Runion 1968, p. 6 of Action Plan - 

Part 2). The laundry room also contained a CAM (Clark 1968b). 

A23.0 LOAD-IN FACILITY

The Load-in Facility (LIF) was built by the WVDP adjacent to the west wall of the EDR. It was built to 

make use of the EDR as the primary access for moving canisters and replacement equipment into the

Vitrification Cell or the CPC (WVDP 2005). 

The dose rate in the LIF is less than 0.1 mrem/hr (WVDP 2005). 

A24.0 LOW LEVEL WASTE TREATMENT FACILITY

The LLWT facility began operating on May 24, 1971. It utilized ion exchange methods to remove 

cesium and strontium from liquid wastes collected by the Interceptor system. The effluent from the 

LLWT facility was stored in Lagoons 4 or 5. It would then be analyzed and transferred to Lagoon 3 for 

discharge to surface streams. If the analyses of wastes in Lagoons 4 or 5 exceeded concentration 

limits, the waste was recycled back to Lagoon 2. NFS (1976) says the LLWT facility removed “greater 

than 95% of the cesium and strontium” from liquid wastes. 

A24.1 RADIOLOGICAL CONDITIONS

A24.1.1

The area around the LLWT facility’s ion exchange beds read 30 mR/hr at contact and 5 mR/hr in the 

aisle shortly after it began operating in May 1971. Exposure rates from the first drums of sludge were 

50 mR/hr contact 5 mR/hr at 3 ft (Wenstrand 1971d). 

External Exposure Data

A25.0 LOW LEVEL WASTEWATER TREATMENT BUILDING

The Low Level Wastewater Treatment (LLW2) Building was built by the WVDP and put into service in 

1998. Its purpose is to remove contaminants (90Sr and 137Cs primarily) from radioactive wastewater

via ion exchange and filtration. It treats water from Lagoon 2 and groundwater from wells on the 

North Plateau (WVDP 2005). 

Spent resin from the ion exchange modules is dewatered in a packaging area in the northeast end of 

the building. The liquid from the dewatering operation goes to a sump to be processed as 

wastewater. The spent resin is packaged and transferred out for storage and disposal (WVDP 2005). 

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A25.1 RADIOLOGICAL CONDITIONS

A25.1.1

Survey data show contact readings on one of the ion exchange modules up to 11 mrem/hr. A 2003 

survey of the sump exit piping showed 4 mrem/hr. General area dose rates were less than 1 mrem/hr 

in 2005 (WVDP 2005). 

External Exposure Data

A25.1.2

Most removable contamination surveys do not show any detectable activity. A smear of the sump 

grating cover showed 1,815 dpm/100 cm2 removable beta in 2003 (WVDP 2005). 

Removable Contamination Data

A26.0 LOWER EXTRACTION AISLE

The LXA was immediately north of the extraction cells on the 114.5-ft elevation of the plant. It 

contained two PSCs with gloveboxes. The LXA was an operations and maintenance area that

provided access to piping and instrumentation penetrations into the extraction cells and the UPC. 

A26.1 RADIOLOGICAL CONDITIONS

As with numerous areas in the process building, the ventilation duct passing through the LXA was a 

chronic source of external personnel exposure. 

A26.1.1

The July 27, 1972, summary of plant decontamination activities says the principal radiation sources in 

the LXA were the ventilation duct and ventilation washer (Wenstrand 1972d, p. 3). 

External Exposure Data

In early 1968 the LXA had an area of elevated exposure rates (in excess of 100 mR/hr) due to backup 

in a steam supply line that supplied a jet in one of the process cells. Shielding was supposed to have 

been installed on the line by March 1, 1968 (Runion 1968, p. 11 of Action Plan - Part 2). 

The exposure rate under the ventilation duct in the LXA was 200 mR/hr before it being flushed on 

October 3, 1969. 

As of 2004 the general area dose rate in the LXA was typically 0.1 mrem/hr with higher levels present 

near ventilation ducts (WVDP 2005). 

A27.0 LOWER WARM AISLE

The LWA was immediately adjacent to the extraction cells on the ground level. It contained shielded 

niches (concrete enclosures) for equipment such as air-operated pumps, valves, etc. that serviced the 

extraction cells. The LWA would become contaminated during replacement or repair of this 

equipment and from movement of contaminated filters from the OGBR through the LWA to the waste 

truck for disposal (Runion 1968, p. 9 of Action Plan - Part 2). 

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General decontamination was performed on the LWA between 1972 and 1974. Additional cleanup 

and modifications (to support the WVDP) occurred in the early 1980s (WVDP 2005). 

A27.1 RADIOLOGICAL CONDITIONS

A27.1.1

On November 7, 1967, a “leaking drain pipe” on the west wall of the LWA read 5 R/hr at 6 in. 

(Wenstrand 1967a). 

External Exposure Data

DOG filters being stored in the LWA on March 29, 1968, were causing exposure rates from 1 R/hr to 

50 R/hr in the aisle (Wenstrand 1968b). 

In 1987 the dose rate in LWA Niche 1 ranged up to 2,700 mrem/hr. Rates for Niche 2 were up to 758 

mrem/hr and up to 96 mrem/hr in Niche 3 (WVDP 2005). 

Typical dose rates in the LWA (not in the niches) as of 2005 are 0.1 mrem/hr to 0.3 mrem/hr (WVDP 

2005).

A27.1.2

Removable beta contamination in the LWA was 100,000 dpm/100 cm2 to 600,000 dpm/100 cm2 on 

November 9, 1967. Removable alpha was up to 18,000 dpm/100 cm2

. 

Removable Contamination Data

On November 23, 1967, the LWA showed maximum levels of removable contamination on the floor of 

750,000 dpm/100 cm2

 beta and 140,000 dpm/100 cm2 alpha. Horizontal surfaces showed removable 

up to 1.8E+06 dpm/100 cm2 beta and 4.5E+06 dpm/100 cm2 alpha. 

Removable contamination in the Product Packing and Shipping and LWA airlock ranged from 18,000 

dpm/100 cm2 to 6E+06 dpm/100 cm2 beta on January 16, 1968. On January 20, 1968, the LWA 

showed removable levels in equipment niches of 500,000 dpm/100 cm2 to 9E+06 dpm/100 cm2 beta 

and 1E+06 dpm/100 cm2 to 9E+06 dpm/100 cm2 alpha (Wenstrand 1968f, p. 2). 

In May 1968 “aggressive” decontamination efforts reduced removable beta contamination levels to 

Zone 3 limits in both the UWA and LWA. Removable alpha contamination in the LWA was still at 

Zone 4 levels (Keely 1968e).

A spill on June 30, 1969, resulted in removable contamination levels in the LWA of 1E+06 dpm/100 

cm2

 (Keely 1969d). 

A27.1.3

In April 1968 the Plutonium Cell and the LWA had to be “fogged” with water to reduce airborne 

radioactivity levels to less than 1000 MPC so entry could be made using supplied air (Keely 1968j). 

The excessive airborne concentration in the LWA resulted from a filter change on April 4, 1968 

(Author unknown 1968, p. 2). 

Airborne Contamination Data

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A28.0 MAIN OFFICE BUILDING

The main Office Building served as the main entrance to the processing plant. It had three floors and 

was located west of the CPC. Between the Office Building and the CPC was the Chemical Viewing 

Aisle. 

The first floor of the Office Building included the Hot Lobby and locker room areas. The Hot Lobby 

was a dressout area where workers donned basic protective clothing. It served as a buffer area 

between the Office Building and the processing plant. The office building had another lobby area that

was differentiated from the Hot Lobby, i.e., they were separate areas. 

A28.1 RADIOLOGICAL CONDITIONS

The ventilation and acid recovery systems were chronic sources of external exposure in the office 

building. Contamination in the Southwest Stairwell was also a contributor. 

A28.1.1

Radiation levels in the third floor offices increased to 5 mR/hr in September 1966. Personnel were 

moved to the Annex Building to decrease their exposure (Keely 1966). 

External Exposure Data

“Background in the Control Room and Office areas sometimes rose to 10 mR/hr” from activity in the 

recovered acid storage tanks in the Hot Acid Cell (NFS 1976). 

The July 27, 1972, summary of plant decontamination activities says radioactive piping was removed 

from the northwest corner of the OGA, resulting in “some reduction” in radiation levels in the 3rd floor 

office building. The report also says radiation levels in the Office Building originating from the 

Southwest Stairwell have not changed from a previous survey in March 1972 (Wenstrand 1972d, p. 

2). 

Most areas in the Office Building showed dose rates less than 32 μrem/hr in April and May of 2005. A 

maximum of 1.4 mrem/hr was measured on the first floor (WVDP 2005). 

A28.1.2

The OGA and Southwest Stairwell were contaminated to Zone 4 levels by an acid leak in the OGA on 

October 23, 1969. Contamination was spread to the Office Building and Chemical Viewing Aisle 

(Wenstrand 1969b). The spill also contaminated an area beneath the floor of the men’s shower room 

on the ground level (WVDP 2005). 

Removable Contamination Data

A29.0 MAINTENANCE SHOP AND WAREHOUSE

These facilities were separate from the process building. Maintenance or repair operations on 

contaminated components were sometimes carried out in the Maintenance Shop. However, this was 

not routinely a radiological area. No indication of radioactive material being present in the warehouse 

has been found in references reviewed to date. 

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A29.1 RADIOLOGICAL CONDITIONS

A temporary contamination zone was established in the Maintenance Shop in March 1968 for the 

repair of a PAR Manipulator (Wenstrand 1968b). 

An AEC inspector looked at the “new” Maintenance Building and the Warehouse during an October 

1973 inspection. No RAM was found in the warehouse. Two fuel canisters were found in the 

maintenance building, one of which appeared to have been modified by having a new section welded 

on. The contact exposure rate was 1 mR/hr (gross). Smear surveys and nasal swabs had been 

collected during the welding (Nelson 1973b, p. 2 of enclosure 2). 

A30.0 MANIPULATOR REP AIR ROOM

The MRR was located north of the MOA and below the PMCR on the ground (100-ft) level of the 

plant. A ceiling hatch provided access to the PMCR for repair and adjustment of the power 

manipulator used in the PMC. The manipulator would be positioned in the Crane Room so its hand, 

wrist, and arm could be extended through the hatch into the MRR. This allowed it to be repaired 

without workers being directly exposed to the radiation in the Crane Room or from the manipulator’s 

bridge. 

A30.1 RADIOLOGICAL CONDITIONS

A30.1.1

Exposure rates in the MRR were significant during operations. The shielding of the MRR was found 

to be inadequate, causing exposure rates up to 50 mR/hr in adjacent areas. 

External Exposure Data

In 2003 the dose rate in the MRR ranged from 10 mrem/hr to 12 mrem/hr (WVDP 2005).

A30.1.2

On February 13, 1968, the MRR airlock showed removable beta contamination ranging from 76,000 

dpm/100 cm2 to 900,000 dpm/100 cm2

 (Wenstrand 1968e, p. 2). 

Removable Contamination Data

Removable contamination up to 4500 dpm/100 cm2 alpha and 112,500 dpm/100 cm2 beta were 

measured in the MRR in 2003 (WVDP 2005). 

A31.0 MECHANICAL OPERATING AISLE (NORTH, EAST AND WEST)

Head-end mechanical operations were directed from operating aisles adjacent to the shielded viewing 

windows of the hot cells. The MOA surrounded three sides of the PMC on the ground (100 ft) level of 

the plant. The two sections of the MOA that ran adjacent to the east and west sides of the PMC were 

therefore often referred to independently as the East Mechanical Operating Aisle (EMOA) and the 

West Mechanical Operating Aisle (WMOA). Occasionally there will also be a reference to the north 

MOA, usually in relation to the MSM Repair Shop that was north of the PMC (i.e., on the other side of 

the MOA). 

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A31.1 RADIOLOGICAL CONDITIONS

A31.1.1

An inspection by the plant housekeeping committee on January 27, 1967, found the plant to be in a 

“deplorable” condition. Radiological items noted in the inspection report include smoking in the 

WMOA, which apparently was not prohibited at the time, and an MSM in the aisle with fingers reading 

5 R/hr contact and 10 mR/hr at 6 ft. The report says the MSM had been there for some time, and 

should either be decontaminated or moved to a less-traveled area. Shear filters in the EMOA were 

reading 10 mR/hr through the concrete wall at 10 in. This radiation area was not posted. 

External Exposure Data

Air cylinders used in the PMC to operate various remote tools exhausted into a header that was 

located in the MOA approximately 10 ft above the floor. Activity accumulating in the header resulted 

in contact readings of 10 mR/hr to 300 mR/hr. General area exposure rates in the MOA were said to 

have been “less than 100 mR/hr” (Runion 1968, p. 3 of Action Plan - Part 2). In March 1968 filters 

were installed in pneumatic piping in the PMC, reducing exposure rates in the MOA to where only a 

“small area” was still designated a radiation area (Clark 1968d). A July 1968 report said the filters 

prevented all but a 3-ft section of the piping from being a radiation area. The 3-ft section was shielded 

so that the exposure rate at 18 in. was “typically 10 mR/hr” (Clark 1968a, p. 2). NFS (1976) states 

buildup of radioactivity in the plant air exhaust lines from cylinders in the PMC caused exposure rates 

up to 24 mR/hr next to the WMOA windows. 

The MRR was not adequately shielded and exposure rates in the North Mechanical Operating Aisle 

“reached 50 mR/hr on occasions.”

A July 27, 1972, summary of plant decontamination activities mentions removal of contaminated 

piping and other sources from the WMOA and cleanup of fixed contamination on floors and walls in 

the EMOA. The report says the filter units in operation on the west end of the FRS were causing 

“significant dose rates” in the EMOA. It also says decontamination efforts in the MRR had resulted in 

a significant reduction of dose rates in the EMOA (Wenstrand 1972d, p. 1). Another report says the 

decontamination efforts in the MRR reduced the exposure rate in the MOA by a factor of four 

(Wenstrand 1972j). 

Typical dose rates in the MOA in 2005 range from 0.1 mrem/hr to 5.4 mrem/hr (WVDP 2005). 

A31.1.2

On October 24, 1969, at 5:00 pm the EMOA and WMOA, along with numerous other areas of the 

plant, became contaminated to Zone III levels. The MOAs were contaminated to 2000 to 3000 

dpm/100 cm2 beta. Maintenance work was postulated as the cause of the widespread tracking of 

contamination, but no definite cause was established. The report mentions that high backgrounds 

make detection of “low” levels of contamination (800 to 1000 dpm beta) difficult. Overall, the cause 

was attributed to either failure of persons to self-monitor (using hand and foot monitors) or 

ineffectiveness of these monitors (Jaeger ca. 1969). 

Removable Contamination Data

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A31.1.3

In December 1966 airborne alpha activity in the MOA rose above NFS’ concentration limit of 4E-11 

μCi/cm3 for canister masks. The activity was attributed to contamination in the CAA that resulted from 

maintenance work in the Plutonium Cell (Keely 1967e).

Airborne Contamination Data

A report dated July 8, 1970, states the east and west MOAs are some of the areas that show air 

concentrations greater than “40 MPC-hr” [sic] during duct flushing operations (TKW 1970). 

A32.0 MSM SHOP/CONTACT SIZE REDUCTION FACILITY

Construction of a new MSM Repair and Decontamination shop (to replace the Hot Shop) was under 

way in May 1968 (Clark 1968d). It was completed within 1 year. The MSM Shop was used for the 

repair of MSMs from the mechanical and analytical cells. The MSMs were decontaminated remotely 

while still in place to reduce the exposure rate to about 1 R/hr. The MSMs were then packaged and 

transferred to the MSM Shop for further decontamination and repair. 

The MSM Shop was decontaminated and renovated between July 1982 and June 1983. 

Contamination had penetrated the original floor to the extent that it had to be removed and repoured. 

As of 1994 the MSM Shop was used to repair and adjust 25 to 30 manipulators a year. The 

manipulators were first decontaminated in an area adjacent to the MSM Shop. 

The MSM Repair Shop transitioned to the Contact Size Reduction Facility beginning in 1987. 

A32.1 RADIOLOGICAL CONDITIONS

A32.1.1

Exposure rates up to 50 mR/hr were present in the MSM Shop from activity in the adjacent MRR. 

External Exposure Data

A July 27, 1972, summary of plant decontamination activities said the “background” in the manipulator 

repair area of the MSM Shop originated from the HEV System filters, the demister and the duct 

located on the east wall. It stated the “small room in the southwest corner of the MSM Shop” 

contained several sources that “should be packaged and removed ... or buried”, and that the drain line 

from the PMCR needed to be shielded. It said radiation levels in the MSM Shop originating from the 

MRR had been reduced, but other sources had been added, and that the north MSM Shop contained 

a “high radiation source” located by the double doors that needed to be packaged for storage or 

shielded (Wenstrand 1972d, p. 2).

A32.1.2

Removable contamination levels exceeded 50,000 dpm/100 cm2 in the “new” MSM Repair Area in 

May 1969 (Keely 1969c, p. 1). 

Removable Contamination Data

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A33.0 OFF GAS AISLE AND PROCESS SAMPLE CELL 3

The OGA was in the southwest area of the process building on the 131-ft elevation. It was also 

known as the Off Gas Operating Aisle or the Acid Recovery Aisle. The Acid Recovery Cell was in the 

southwest corner of the process building on the same level. 

The OGA was used for monitoring and controlling operations in the OGC, the Acid Recovery Cell, and 

part of the CPC. It also provided access to PSC-3 and acted as a passageway between laboratories, 

the Office Building and the south and southwest stairs (WVDP 2005). PSC-3 contained three 

samplers for fission product solutions in XC1 (Runion 1968, p. 16 of Action Plan - Part 2).

A33.1 RADIOLOGICAL CONDITIONS

The OGA was impacted by the substantial spill of contaminated acid from the Acid Recovery Cell in 

1967 (see Section A1.0). 

A33.1.1

On November 12, 1967, the exposure rate in PSC-3 was 70 mR/hr to 150 mR/hr. The west side of 

the glove box read 11 rad/hr. 

External Exposure Data

Radioactivity in the acid recovery system created significant exposure rates in the OGA (Runion 1968, 

p. 16 of Action Plan - Part 2; NFS 1976). 

On June 26, 1972, cement blocks were removed from the OGA for burial. The blocks read 5 R/hr 

(general area) (Nelson 1973a, p. 14 of attachment). Dose rates as high as 20 rem/hr were recorded 

in the OGA in 1972. Concrete was removed from this area reducing the dose rate to a maximum of 

400 mrem/hr (WVDP 2005). 

The July 27, 1972, summary of plant decontamination activities says the radiation level on the OGA 

roof around the condenser had been reduced by a factor of 10 from decontamination efforts. The 

major source of activity then became the backup filters (Wenstrand 1972d, p. 3). 

Work was performed in the OGA the week of February 20, 1974, to remove a contaminated stainless 

steel pipe embedded in the floor (O’Reilly 1975). The general area exposure rate at the start of the 

work was 15 mR/hr to 20 mR/hr. The exposure rate in contact with the pipe was 400 mR/hr.

As of 2005 dose rates in the OGA generally range from 0.1 mrem/hr to 0.7 mrem/hr (WVDP 2005). 

A33.1.2

Smearable contamination reached Zone 4 levels due to the leak in the DOG vent pipe in January 

1969 (Keely 1969h, p. 2).

Removable Contamination Data

PSC-3 became contaminated from leakage of recovered acid from a valve in a sampler. 

Contamination also resulted from penetrations in the floor and airflow into the area from the OGA. 

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Staff stated the blocks removed from the OGA on June 26, 1972, were grossly contaminated, having 

been used to shield an acid recovery line. Analysis showed the contamination on the blocks included 

106Ru, 90Sr, 239Pu, and 95Zr/Nb (Nelson 1973a, p. 14 of attachment). 

During decontamination work the week of February 20, 1974, removable levels were generally 4000 

dpm/100 cm2 to 5000 dpm/100 cm2 with a maximum of 200,000 dpm/100 cm2

. Potential nuclides in 

the contamination included mixed fission products, uranium, and plutonium (O’Reilly 1975). The 

report does not say whether the stated removable levels were alpha or beta. 

A33.1.3

In January 1969 airborne activity in the OGA reached 20 MPC following failure of a temporary repair 

of the DOG vent pipe. 

Airborne Contamination Data

A34.0 OFF GAS BLOWER ROOM

The OGBR was in the southwest section of the plant on the ground level, directly north of the ARPR. 

It contained blowers, filters, and scrubber recirculation pumps for the process off gas system. 

The OGBR was decontaminated between November 1975 and June 23, 1976. Equipment (blowers 

and filters) was removed and new concrete was poured to repair parts of the floor that had been 

eroded away by spills from the ARC above (WVNSC 1994). 

The OGBR was entered several times in the 1980s to replace HEPA filters. As of 1994 the OGBR 

had not been entered since the fall of 1992 (WVNSC 1994). 

A34.1 RADIOLOGICAL CONDITIONS

A34.1.1

A survey inside the 6 G-3 scrubber pump niche in the OGBR on January 4, 1968, showed a wholebody exposure rate of 4 R/hr at 18 in. The extremity dose rates ranged from 18 rad/hr to 20 rad/hr 

(Wenstrand 1968f, p. 2). 

External Exposure Data

On January 15, 1968, the DOG filter in the OGBR read 600 R/hr on contact and 100 R/hr at 12 in. It 

read 1000 rad/hr at 10 in. (Wenstrand 1968f, p. 2). Changing of the DOG and VOG filters was a 

contact maintenance activity. 

As of early 1968 an operator had to enter the OGBR once a day for operation of manual valves to 

change solutions in the off gas scrubbers. The exposure rate in this area was 300 mR/hr (Runion 

1968, p. 18 of Action Plan - Part 2). 

In April 1968 “easily” removable contamination continued to be a problem in the OGBR (as well as the 

ARPR), giving skin dose rates of 120 mrad/hr (Keely 1968j). 

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In May 1968 a TLD disk was used to measure the dose rate from the floor of the OGBR, giving a 

result of 165 rad/hr (Keely 1968e). 

In June 1968 the typical working area exposure rate in the OGBR was 1 to 2 R/hr (Keely 1968a). 

Exposure rates of 15 R/hr were encountered during cleanup of the “niche” in the OGBR in November 

1972 (Morrow 1972d). 

HEPA filters removed for replacement in 2000 and 2001 read 55 mrem/hr (WVDP 2005). 

In 2003 dose rates in the southwest part of the OGBR ranged from 70 mrem/hr to 700 mrem/hr 

(WVDP 2005).

As of 2005 the general area dose rate in the OGBR ranged from 10 mrem/hr to 80 mrem/hr (WVDP 

2005).

A34.1.2

As of early 1968 the OGBR was chronically contaminated from replacement of the DOG and VOG

filters (Runion 1968, p. 18 of Action Plan - Part 2). 

Removable Contamination Data

In May 1968 cutie pie readings on smears from the OGBR indicated a maximum removable level of 

4E+08 dpm/100 cm2

 (Keely 1968e). 

A35.0 OFF GAS CELL

The OGC was in the southwest section of the plant on the ground level. It was directly north of the 

OGBR. The OGC contained equipment for the VOG and DOG systems. 

A35.1 RADIOLOGICAL CONDITIONS

The OGC was another one of the areas affected by the 1967 acid spill from the Acid Recovery Cell 

above (see Section A1.0). This lead to high radiation levels in the northwest corner of the cell (WVDP 

2005). 

A35.1.1

In August 1972 an entry was made into the OGC for a radiation survey for the first time since fuel 

reprocessing began. The general area exposure rate was between 10 R/hr and 20 R/hr (Morrow 

1972e). The radiation levels along the floor were 20 R/hr to 25 R/hr and as high as 50 R/hr in the 

northwest corner of the cell. 

External Exposure Data

In 2002 dose rates in the area of the DOG scrubber ranged from 94 mrem/hr at 28 ft from the floor to 

1,600 mrem/hr at floor level. Dose rates in the northwest corner of the cell ranged from 270 mrem/hr 

at 28 ft from the floor to 1,500 mrem/hr at the floor. 

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A36.0 PERMANENT VENTILATION SYSTEM BUILDING

The PVS Building was constructed by the WVDP to provide ventilation to the STS Support Building, 

the STS Valve Aisle, and the STS Pipeway. The building contains equipment (filters primarily) to treat 

ventilation exhaust air before release to the atmosphere. The PVS Building has its own stack. 

A36.1 RADIOLOGICAL CONDITIONS

A36.1.1

Exposure rates are reported to be 80 mrem/hr to 170 mrem/hr on contact with the filter housing 

(WVDP 2005). 

External Exposure Data

A36.1.2

Routine surveys do not show detectable removable contamination. Removable beta of 2.5E+06 

dpm/100 cm2 has been measured in the inlet plenum (WVDP 2005). 

Removable Contamination Data

A37.0 PROCESS CHEMICAL ROOM 

The Process Chemical Room (PCR) was above the north end of the CPC on the 144-ft elevation of 

the plant. It was directly north of and adjacent to the Hot Acid Cell. It contained four chemical makeup tanks and associated pumps.

A37.1 RADIOLOGICAL CONDITIONS

A37.1.1

In June 1969 the general exposure rate in the PCR increased from 30 mR/hr to 200 mR/hr with 

localized spots around the sampler reading as high as 5 R/hr (Keely 1969d).

External Exposure Data

The July 27, 1972, summary of plant decontamination activities says radiation levels in the PCR were 

significantly reduced when the Hot Acid Cell was decontaminated.

As of 2005 the dose rate in the PCR was around 1 mrem/hr (WVDP 2005). 

A37.1.2

The July 27, 1972, summary of plant decontamination activities says radiation levels in the PCR were 

significantly reduced when the Hot Acid Cell was decontaminated. The reduced levels allowed the 

detection of contamination in the PCR that had previously been “obscured” by the high background 

(Wenstrand 1972d, p. 3). 

Removable Contamination Data

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A38.0 PROCESS MECHANICAL CELL

The PMC is where extraneous hardware was removed from fuel assemblies, which were then 

sheared into small pieces for dissolving. The sheared fuel dropped through a discharge chute into 

baskets in the GPC below. 

The PMC included high-workload manipulators for handling fuel elements. Maintenance and repair of 

these manipulators (performed after they were removed from the cell) was a significant source of 

personnel exposures because of the substantial contamination from fuel particles. Exhaust from air 

cylinders used in the PMC to operate various remote tools was a source of contamination and 

elevated exposure rates in other areas of the plant, such as the MOA. 

A general cleanup of the PMC was conducted between October 22, 1971, and September 15, 1972 

(WVNSC 1994). The WVDP completed deactivation of the PMC in 2004 (WVDP 2005). 

A38.1 RADIOLOGICAL CONDITIONS

The PMC was never a routinely occupied area. 

A survey of the PMC in April 1972 showed exposure rates ranging from 1,500 R/hr to 4,500 R/hr 

(Wenstrand 1972j). 

In 1986 gamma exposure rates in the PMC ranged from 0.8 R/hr to 40 R/hr at 6 ft from the floor. 

Beta-gamma levels ranged from 40 rem/hr to 240 rem/hr (WVDP 2005).

Surveys performed after the 2004 cleanup in grid locations 3 ft above the floor showed dose rates 

ranging from 1.5 rem/hr to 40 rem/hr (WVDP 2005). 

A39.0 PRODUCT P ACKAGING AND SHIPPING, PRODUCT PACKAGING AND HANDLING, 

URANIUM PRODUCT CELL, AND URANIUM LOADOUT AREAS

The aqueous plutonium and uranium nitrate products from fuel reprocessing were packaged in 10 L 

shipping containers in the PPH area. They were then shipped by truck from the PPS area above. 

The two were connected via an airlock (WVDP 2005). During operations the PPH contained glove 

boxes used for filling of product bottles. One was used for plutonium and the other for uranium.

The PPH became the Waste Reduction and Packaging Area under the West Valley Demonstration 

Project (WVNSC 1994). In 2005 it was being used for compressing LLW into metal storage 

containers using a 50-ton hydraulic compactor (WVDP 2005). 

The UPC was on the ground level directly north of the PPC. It contained tanks for the collection and 

storage of uranyl nitrate before analysis and shipment. The function of these tanks was modified in 

1988 to where they became hold tanks for the WVDP Low-level Waste Treatment System (WVDP 

2005). 

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The ULO, also on the ground level, was directly east of the UPC. The ULO contained a 4,000-gal

weigh tank used to measure shipments of uranyl nitrate before transfer to tanker trucks or to the back 

of the cell for further processing (WVDP 2005). 

A39.1 RADIOLOGICAL CONDITIONS

A39.1.1

May 1966: Several shipments of uranium product were made and two bottles of plutonium product 

were loaded out. The maximum radiation level on the Pu bottles was 130 mR/hr (Loud 1966d). 

External Exposure Data

Exposure rates at 6 in. from the Product Packaging and Handling loading dock on January 20, 1968, 

were 2 R/hr whole body and 17 rad/hr extremity (Wenstrand 1968f, p. 2). 

In 1996 dose rates in the UPC ranged from 10 mrem/hr to 450 mrem/hr. Surveys in 2003 showed

dose rates of 120 mrem/hr on the underside of one of the waste tanks and 110 mrem/hr between the 

two tanks (WVDP 2005). 

As of 2004 the dose rate in the PPH area was less than 1 mrem/hr (WVDP 2005). 

As of 2005 the dose rate in the ULO area was typically less than 1 mrem/hr, with levels up to 29 

mrem/hr in the pump room (WVDP 2005). 

A39.1.2

Decontamination efforts in the UPC in April 1972 reduced the typical removable alpha contamination 

levels from 107 dpm/100 cm2 to 105 dpm/100 cm2 (Wenstrand 1972j). 

Removable Contamination Data

In 1996 surveys showed removable levels up to 32,000 dpm/100 cm2 alpha and 348 dpm/100 cm2

beta in an area north of one of the tanks (WVDP 2005). 

Low levels of removable contamination were present in the ULO area as of April 2005. Removable 

was not detected elsewhere in the area (WVDP 2005). 

A39.1.3

The airborne activity levels in the PPS area exceeded the MPC during the week of March 18, 1968 

(Wenstrand 1968b). 

Airborne Contamination Data

In June 1969 long-lived alpha air activity in the PPH area exceeded the MPC for “a major portion of 

the month.” The air activity was said to originate from contamination present on the north wall by the 

High Enriched Loadout Box (Keely 1969d). 

A40.0 PULSER EQUIPMENT AISLE

The PEA was immediately south of the XCR on the 160-ft elevation of the process building. The PEA 

included air pulsers for the extraction columns in the extraction cells and other equipment. This 

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equipment was removed in 1980 and the area was reconfigured to support operation of the LLW

treatment system. The former PEA no longer exists (WVDP 2005). 

A40.1 RADIOLOGICAL CONDITIONS

A40.1.1

The airborne activity levels in the PEA exceeded the MPC during the week of March 18, 1968. The 

same document states this condition existed in both the PEA and the PPS area, suggesting the MPC 

of concern was that for alpha (239Pu) (Wenstrand 1968b, p. 1). 

Airborne Contamination Data

A41.0 REMOTE HANDLED WASTE FACILITY

The RHWF was built by the WVDP. It went into operation in June of 2004. It is used for size 

reduction and packaging of equipment containing large amounts of radioactive materials. 

A42.0 SAMPLE STORAGE CELL, ANALYTICAL HOT CELLS, AND ANALYTICAL 

DECONTAMINATION AISLE 

The Sample Storage Cell (SSC) was on the east side of the CPC at the 131-ft elevation (i.e., the 

analytical level of the plant). It contained a conveyor-elevator for transferring samples to the 

Analytical Hot Cells or laboratories. The SSC had three shielded viewing windows and three sets of 

manipulators in its north wall. As of 2005 the SSC remained highly contaminated and showed dose 

rates from 126 mrem/hr up to nearly 20 rem/hr (WVDP 2005, p. C-49). 

Entries were sometimes made into the SSC for periodic maintenance of equipment in the Analytical 

Hot Cells that could not be removed. These entries were made via the Analytical Decontamination 

Aisle (ADA). The SSC would be remotely decontaminated before these entries, and internally 

contaminated items such as sample bottles would be removed via the conveyor system. 

On the east side of the ADA were five Analytical Hot Cells and Sample Cell 2C. The ADA was 

sometimes referred to as the Analytical Cell Decontamination Area. 

A43.0 SCRAP REMOVAL AREA

The SR area was on the ground (100 ft) level of the process building northeast of the CPC. This area 

is also sometimes referred to as the SR Cell or SR Room. 

Leached hulls and process wastes (including failed equipment) were placed in 30-gal drums for burial. 

The drums were loaded inside the GPC, then passed through a hatch to the SR area. They were 

sprayed with water as they passed through the hatch and then transferred to a shielded cask inside 

the SR area. Excess water was allowed to drip off before the drums were placed in the transfer cask. 

The cask was placed on a trailer (which had been backed into the SR area) and both were 

decontaminated before being released to go to the HLW burial area.

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Decontamination of the SR area took place in 1972 and again in 1980. D&D of the SR Room took 

place between March and September 1986. The SR Room now contains Tank 7D-13, which formerly 

received liquids from the plant laundry and analytical laboratory drains. As of 1994 the tank received 

liquids from laboratory drains and from drum flushing operations at the Concrete Solidification System 

(WVNSC 1994). 

A43.1 RADIOLOGICAL CONDITIONS

A43.1.1

A dissolver removed from the CPC in July 1969 read 30 R/hr at contact with its shield cask and 50 

mR/hr at 50 ft (upon removal from the SR area) (Keely 1969e).

External Exposure Data

The 1986 D&D of the SR area reportedly reduced exposure rates in the SR Room to less than 1 

mR/hr (WVNSC 1994). However, other references report levels from 1 mrem/hr to 40 mrem/hr at 1 ft

from the floor in 2003, and up to 10 mrem/hr at 3 ft from the floor (WVDP 2005). 

A43.1.2

The SR area was a chronic source of contamination because of the waste packing and removal 

operations. Typical contamination levels for the drums is said to have been in excess of 500,000 

dpm/100 cm2 as they came from the GPC to the SR area (Runion 1968, p. 4 of Action Plan - Part 2). 

Removable Contamination Data

A43.1.3

In January 1967 repair of the seal around the SR door resulted in differential pressure between the 

GPC and the SR Area, with the latter becoming an airborne activity area on several occasions

thereafter (Keely 1967d, p. 2). 

Airborne Contamination Data

A44.0 SOLVENT STORAGE TERRACE 

The Solvent Storage Terrace (SST) was located on the south side of the process building on the roof 

of the UWA at the 131-ft elevation. Storage tanks for recovered acid and solvent wastes were located 

on the SST. The Recovered Acid Storage Tank received acid from the bottom of the acid fractionator 

located in the Acid Recovery Cell. The solvent waste tanks received solvent from the extraction cells 

(WVNSC 1994). 

A44.1 RADIOLOGICAL CONDITIONS

The SST became contaminated from spills and leaks from the tanks. Liquids from the SST drained to 

an unlined holding pond near Lagoon 1 known as the “solvent dike” (WVNSC 1994). 

A44.1.1

On November 4, 1967, the solvent storage tank read 2.5 R/hr to 3 R/hr on contact. 

External Exposure Data

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In June 1968 a survey of the SST showed hot spots on the floor reading up to 100 R/hr. The 

exposure rate at the bottom of the solvent storage tank was 300 R/hr (Keely 1968a). 

As of 2005 dose rates on the SST were typically 0.1 mrem/hr or less (WVDP 2005). 

A44.1.2

In July 1968 smears from the SST indicated removable beta contamination levels of 1E+08 dpm/100 

cm2

 (Keely 1968k). 

Removable Contamination Data

In March 1972 a leaking connection during solvent loadout caused removable beta contamination up 

to 107 dpm/100 cm2 on the SST (Wenstrand 1972k). 

A45.0 SUPERNATANT TREATMENT SYSTEM SUPPORT BUILDING

The STS Support Building was built by the WVDP to support STS operations. It is a two-story 

building located adjacent to and on top of the HLW storage tank 8D-1. The first level contains the 

valve aisle, operating aisle, and control room. The second story houses tanks and delivery systems. 

The STS Support Building supported STS operations from 1988 through 1995. Subsequently it was 

used to treat vitrification waste streams and for processing of sodium-bearing wastewater. 

A45.1 RADIOLOGICAL CONDITIONS

A45.1.1

In May of 2002 dose rates in the valve aisle ranged from 0.9 rem/hr to 8.2 rem/hr 4 ft from the back 

wall. The dose rate ranged from 0.5 rem/hr to 3.3 rem/hr 12 in. from the viewing windows (WVDP 

2005). 

External Exposure Data

A45.1.2

Surveys on the first and second floors outside the valve aisle typically do not show detectable 

removable contamination. Removable contamination is said to be “very high” inside the valve aisle 

(WVDP 2005). 

Removable Contamination Data

A46.0 UPPER EXTRACTION AISLE

The UXA was immediately north of the extraction cells on the 144-ft elevation of the process building. 

It served as an operations and maintenance area for extraction cell and other equipment. The area 

included valves, piping, instrumentation, ventilation ducts, and stack monitoring equipment. 

A46.1 RADIOLOGICAL CONDITIONS

A46.1.1

As of early 1968 a high radiation area existed on the UWA in the area of a recovered acid line (Runion 

1968, p. 11 of Action Plan - Part 2). 

External Exposure Data

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Radiation levels around the 7C-1 jet in the UXA reached 10 R/hr at 18 in. twice during the month of 

May 1969 (Keely 1969c, p. 2).

The July 27, 1972, report on plant decontamination status said radiation from the west wall of the UXA 

appeared to be caused by “a shielding flaw between the aisle and the steam ventilation filter bank” 

(Wenstrand 1972d, p. 3).

In 2004 the general area dose rate in the UXA ranged from 0.1 mrem/hr to 6 mrem/hr. In 2005 the 

range was 0.1 mrem/hr to 4.0 mrem/hr (WVDP 2005). 

A46.1.2

Horizontal surfaces in the UXA showed removable levels up to 280,000 dpm/100 cm2 beta on January 

1, 1968 (Wenstrand 1968f, p. 3). 

Removable Contamination Data

A jet was mistakenly left open on May 25, 1972, allowing an estimated 1.5 to 2 gal of contaminated 

acid and water to leak onto the floor of the UXA. Surveys of the surrounding floor area showed 

50,000 cpm alpha. The activity in the puddle was determined to be 239Pu. 

A47.0 UPPER WARM AISLE AND PROCESS S AMPLE CELL 1 

The UWA was immediately above the LWA, adjacent to the extraction cells on the 114.5-ft elevation. 

Like the LWA, it contained concrete niches that housed pumps, valves, and other equipment used in 

the solvent extraction process. There were six of these niches extending into the UWA. The 

equipment in the niches required routine maintenance or replacement, resulting in contamination of 

the area (Runion 1968, p. 10 of Action Plan - Part 2).

The UWA also contained PSC-1, which was on the northeast end of the aisle. PSC-1 was a shielded 

stainless-steel box that housed a glovebox and samplers for sampling purified plutonium and uranium 

solutions from the PPC. 

NFS decontaminated the UWA, including the niches, in the 1972 to 1974 period. They also replaced 

several of the pumps. The aisle and niches were also decontaminated by the WVDP in 1985 (WVDP 

2005). Equipment niches and associated equipment were removed from the UWA in 1986 (WVNSC 

1994). 

A47.1 RADIOLOGICAL CONDITIONS

A47.1.1

A September 18, 1967, letter from NFS to the AEC says the UWA area in general (i.e., outside the 

pump niches) only became a high radiation area when maintenance work was being performed 

(Runion 1967a, p. 2). 

External Exposure Data

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On November 13, 1967, the exposure rate at the top of the 13G-2 pump niche in the UWA was 2.5 

R/hr to 3.0 R/hr. The pump read 60 rad/hr and 12 R/hr at 2 in. The 13G-1 pump niche read 6 R/hr at 

the top and 20 R/hr to 30 R/hr at the bottom. The pump stand read 50 R/hr at contact. 

Surveys in the UWA on January 20, 1968, showed exposure rates of 10 R/hr and dose rates up to 35 

rad/hr (Wenstrand 1968f, p. 2). 

Typical dose rates in the UWA as of 2005 ranged from 0.1 mrem/hr to 0.5 mrem/hr for the general 

area. There was a hot spot up to 35 mrem/hr on a drain line and up to 15 mrem/hr on the south wall. 

Surveys inside pump niches 1 through 5 in December 2003 to January 2004 showed the following 

maximum levels, each near the floor (WVDP 2005).

• UWA pump niche 1 = 40 mrem/hr

• UWA pump niche 2 = 1.8 mrem/hr

• UWA pump niche 3 = 45 mrem/hr

• UWA pump niche 4 = <1 mrem/hr

• UWA pump niche 5 = <1 mrem/hr

A47.1.2

On November 17, 1967, removable alpha in PSC-1 ranged from 150 dpm/100 cm2 to 3700 dpm/100 

cm2 with hot spots on the floor as high as 320,000 dpm/100 cm2

. 

Removable Contamination Data

Sample bottles in PSC-1 were bagged and transferred into a glovebox. The sampling station would 

periodically become contaminated from these operations, allegedly from inadequate bagging. Entries 

into the PPC also resulted in contamination of the sampling station from airflow reversals into the 

glovebox through its vent. 

Removable beta contamination levels in the UWA ranged from 200,000 dpm/100 cm2 to 1.5E+06 

dpm/100 cm2 on January 22, 1968 (Wenstrand 1968f, p. 3). In February 1968 smears of the UWA 

showed removable beta levels of 1.2E+06 dpm/100 cm2

 (Wenstrand 1968e, p. 2). 

In May 1968 “aggressive” decontamination efforts reduced removable beta contamination levels to 

Zone 3 limits in both the UWA and LWA (Keely 1968e).

In August 1968 removable contamination in the UWA showed a maximum of 8E+06 dpm/100 cm2

(Keely 1968b). 

On October 29, 1968, smearable alpha contamination in PSC-1 reached 250,000 dpm/100 cm2 during 

maintenance work (Keely 1968d). 

Contamination in the UWA was reduced from Zone 4 to low Zone 3 during the month of February 

1969 (Keely 1969f).

In August 1969 removable alpha contamination was 1E+06 dpm/100 cm2 in PSC-1 following 

modifications to sample lines and installation of a glove box (Wenstrand 1969c). 

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Removable contamination levels in the UWA before decontamination in 1985 ranged up to 2,070 

dpm/100 cm2 alpha and 19,800 dpm/100 cm2 beta (WVDP 2005). 

As of 2005 removable levels on the UWA were nondetectable, but levels up to 312,000 dpm/100 cm2

beta-gamma had been measured in the niches (WVDP 2005). 

A47.1.3

In October 1967 air samples “taken on each shift” from the XSA and PSC-1 showed long-lived 

airborne alpha concentrations of 2E-12 μCi/cm3 and 5.4E-11 μCi/cm3

 (Wenstrand 1967d, p. 2). 

Airborne Contamination Levels

A48.0 UTILITY ROOM AND STAIRWELLS

The Utility Room was a concrete block structure adjacent to the process building on its south side. It 

shared its north wall with the process building. This wall was reinforced concrete (WVDP 2005). 

The process building contained four main reinforced concrete stairwells having landings on the 

various levels. These were the

• East Stairs: Near the PPC extending from the ground level to the 160-ft elevation 

• North Stairs: Near the GOA extending from 76.5 ft (below ground level, i.e., the level of the 

GOA) up to the 131-ft (analytical) level

• South Stairs: North of the Utility Room extending from ground level up to the 160-ft elevation

• Southwest Stairs: South of the office area extending from ground level to an elevation of 

152 ft

A48.1 RADIOLOGICAL CONDITIONS

The stairways (South and Southwest in particular), as well as the Utility Room, became contaminated 

from spills and frequent airflow reversals that would occur when doors were opened. These reversals 

resulted in flow of contamination from acid recovery and off gas areas into the stairwells and the Utility 

Room. 

A48.1.1

In June 1969 exposure rates in the Southwest Stairwell increased from 20 mR/hr to 90 mR/hr during 

the “Yankee campaign”; that is, higher burnup fuel (Keely 1969d). 

External Exposure Data

Exposure rates up to 6 R/hr were measured in the lower levels of the Southwest Stairs in the early 

1970s (WVDP 2005). 

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The July 27, 1972, summary of plant decontamination activities says the radiation levels in the shower 

and locker room areas originating from the Southwest Stairwell remained “essentially the same” as 

they did in March 1972 (Wenstrand 1972d, p. 2). 

A typical dose rate of 30 mrem/hr with hot spots up to 125 mrem/hr are reported for the North Stairs 

on the east wall south of a viewing window into the PMCR, but no time frame is given for when or how 

long these exposure rates existed. Levels up to 20 mrem/hr gamma and 130 mrem/hr beta-gamma 

are reported for “different times in other areas” of the north stairwell (WVDP 2005). 

In 1990 the dose rate in the south stairs at the Control Room landing (144-ft level) ranged up to 

9 mrem/hr (WVDP 2005). 

Typical dose rates for the stairwells in 2004 were (WVDP 2005): 

• East stairs = 0.1 mrem/hr to 0.5 mrem/hr

• North stairs = 0.1 mrem/hr to 0.5 mrem/hr.

• South Stairs = 0.1 mrem/hr to 5.0 mrem/hr

• Southwest Stairs = 0.1 mrem/hr to 14 mrem/hr

Typical dose rates for the stairwells in 2005 were (WVDP 2005): 

• East stairs = <0.04 mrem/hr to 1.1 mrem/hr

• North stairs = <0.04 mrem/hr to 0.6 mrem/hr

• South Stairs = 0.2 mrem/hr to 4.8 mrem/hr

• Southwest Stairs = 0.1 mrem/hr to 22 mrem/hr

The elevated dose rates for the Southwest Stairs were at the second landing. 

A48.1.2

In January 1967 gamma analysis on sand from the Utility Room trench, which was reading 20 mR/hr 

to 50 mR/hr, showed “mostly Zr/Nb-95” (Keely 1967d, p. 3). 

Removable Contamination Data

As of early 1968 the South Stairwell had become contaminated from an airflow reversal from the 

OGBR, which was located at the base of the stairway. These reversals occurred whenever the door 

at the base of the stairway was opened (Runion 1968, p. 19 of Action Plan - Part 2). Subsequently, 

the OGA and Southwest Stairway became contaminated since they received air from the South Stairs 

(Runion 1968, p. 20 of Action Plan - Part 2). 

In October 1968 removable contamination on a wall in the Southwest Stairwell was 5E+06 dpm/100 

cm2 around a leak through the wall of the Acid Recovery Cell (Keely 1968d).

The Southwest Stairwell became contaminated to Zone 4 levels by an acid leak in the OGA on 

October 23, 1969 (Wenstrand 1969c). 

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A48.1.3

On July 8, 1970, the East and North stairs showed concentrations “close to” 40 MPC-hr [sic] during 

flushing of the ventilation ducts (TKW 1970). 

Airborne Concentration Data

An air sample collected from the Utility Room in January 1974 showed an alpha concentration of 

5.8E-12 μCi/cm3

. Isotopic analysis of the alpha activity showed the following.

• Pu-238 = 7.6 wt%

• Pu-239 = 76.1 wt%

• Pu-240 = 11.2 wt%

• Pu-241 = 4.3 wt%

• Pu-242 = 0.8 wt%

A49.0 VENTILATION EXHAUST CELL

The VEC was located on the roof of the CPC at an elevation of 148 ft. It contained ductwork, filters, 

blowers, and controls for the main ventilation system. The VEC contained two parallel ventilation 

systems: electric powered for routine operations and steam powered as a backup. The two filter 

banks were therefore identified as “steam” or “electric.” 

A49.1 RADIOLOGICAL CONDITIONS

A49.1.1

During a filter changeout in 1994 the roughing filters read 1.8 rem/hr gamma at 2 in. The HEPA filters 

read 400 mrem/hr to 500 mrem/hr at 2 in. The floor of the filter housing ranged up to 1 rem/hr gamma 

(WVDP 2005). 

External Exposure Data

As of 2005 general dose rates in the VEC ranged from 1 mrem/hr to 100 mrem/hr (WVDP 2005). 

A50.0 VENTILATION SYSTEMS

Overall, the single largest source of personnel exposures at West Valley during its operations era was 

the plant ventilation system. The unanticipated magnitude of the problem of high specific activity fuel 

particles, coupled with plateout of these particles within the system and its failure to maintain 

necessary pressure differentials, was both a direct and indirect source of personnel exposure 

throughout the plant. The main ventilation system serviced the entire reprocessing plant from startup 

until October 25, 1970. On October 25, 1970, a new HEV system came on line that provided 

additional ventilation capacity for the head end process cells and operating aisles. 

A50.1 MAIN VENTILATION SYSTEM

The main ventilation system consisted of an air washer followed by roughing and absolute filters along 

with associated blowers and ducting. After exiting the washer, the ventilation air was routed through a 

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duct along the ceiling of the LXA, around the outside of the laboratories, over the top of the Control 

Room, and into the filter plenum in the VEC. 

Particulate activity collecting on the interior of the ducting gave contact exposure rates from 100 

mR/hr to 800 mR/hr. The system was modified in 1967 to allow periodic flushing of the ducts, which

would offer temporary relief until the contamination built up again. These duct-flushing operations 

would also result in high airborne activity levels throughout the plant. In addition to the chronic local 

and area exposure rate problems associated with the ventilation system ducting, another significant 

source of personnel exposure was that associated with changing the filters (roughing and absolute) in 

the VEC. 

After filtration, the process building ventilation system air was either exhausted through the stack or 

recycled. Recycled air was used in the LLWT Plant, the HLW Storage Facility, and the FRS. The 

recycle air passed through HEPA filters before reuse. 

Other ventilation systems, which served to keep process vessels at negative pressure relative to the 

cells, which were at negative pressure relative to the aisles, included 

• The DOG system

• The VOG system (for process vessels)

• The Waste Tank Farm ventilation system

The DOG system treated gases from the vacuum dissolvers in the CPC and purge gases from the 

fuel bundle shear in the PMC. The DOG system included scrubbers to reduce nitrogen oxides, which 

were also effective for removing particulates and iodines. North and Clark (1968) state silver nitrate 

reactors in the DOG system “have not been put into operation because the long-cooled fuel contains 

an insignificant amount of iodine-131.” The DOG system effluents first passed through HEPA filters 

before discharge via the plant stack. 

The VOG system ventilated gases from process vessels and tanks to the OGC, where it passed 

through scrubber and filters before discharge to the main ventilation system (WVNSC 1994). 

The removal efficiency of the off gas scrubbers was lower than expected (NFS 1976). This resulted in 

higher than expected activity in the DOG and VOG absolute filters. Periodic changing of these filters, 

which was a contact maintenance activity, was a significant source of personnel exposures. 

A document from ca. 1967 gives the criteria for replacement of the DOG filters, saying a filter shall be 

replaced whenever the exposure rate at 2 in. exceeds 500 R/hr. This measurement was to be taken 

whenever the exposure rate through the shield plug exceeded 6 R/hr, i.e., once it got to 6 R/hr the 

plug was to be removed and a measurement taken at 2 in. (Clark 1968e). 

Following the modification of the plant to add the dedicated HEV system (see next section), the 

original building ventilation system serviced the nonprocessing areas of the plant, including the 

operating aisles, office areas, chemical makeup area, maintenance area, FRS pool area and the 

laboratories (NFS 1973b, p. 6.1-8). 

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A50.2 HEAD END VENTILATION SYSTEM

Before 1970, the plant’s main ventilation system served all of the plant, including the head end 

process cells and operating areas. In an effort to relieve the main ventilation system of the substantial 

airborne radioactivity generated in the head end processing, the HEV system was activated in 1970. 

The radioactivity removed by the HEV was greater than expected, resulting in dose rate problems in 

several areas adjacent to the HEV building. In addition, periodic replacement of the HEV filters was a 

source of significant personnel exposures.

The HEV system was added to the plant in response to the operational difficulties encountered from 

the high-activity particulate contamination associated with mechanical segmentation of fuel 

assemblies. The HEV system was expected to increase the differential pressure between occupied 

and contaminated areas, to increase the airflow through contaminated areas where contact 

maintenance was performed (the MSM shop in particular), and to divert the bulk of the particulate 

contamination to a dedicated filter system instead of the main ventilation system filters (Author 

unknown 1969). NFS felt their existing stack sampling and monitoring equipment could handle the 

increased flow. 

The HEV system was licensed on October 20, 1970, and placed into operation October 25, 1970 

(NFS 1971, p. 1). It serviced the cells and operating aisles involved in the mechanical processing of 

fuel. The air flowed into filter trains in the HEV Building and then out to the plant main stack (WVNSC 

1994). 

NFS states the HEV system reduced the exposure rates from the main ventilation system 

components, thus reducing personnel exposures associated with the main ventilation system ducting, 

air washer, etc. Exposure rates were reduced “significantly” in the Control Room, analytical aisles, 

and extraction aisle. Personnel exposures associated with changeout of main ventilation system 

filters were reduced, and the change frequency also decreased. The HEV system also reduced 

contamination levels in the airlocks and crane rooms, and reduced particulate emissions from the 

main stack. However, the location of the HEV exhaust duct near the fuel shearing operation resulted 

in higher than expected loading of the HEV filters and buildup of activity in the duct. The activity in the 

HEV duct and filters was responsible for 69 person-rem/yr of collective dose.

A51.0 VENTILATION WASH ROOM 

The VWR was on the 114.5-ft elevation of the plant, directly south of the PMC. It contained the air 

washer for the main ventilation system. The purpose of the washer was to intercept particulate 

activity from the saw and shear in the PMC. 

A51.1 RADIOLOGICAL CONDITIONS

Buildup of particulates in the ventilation air washer filters after several years of plant operation 

necessitated replacing the filters. Changeout of these filters and the air washer pump were a 

significant source of exposure for maintenance workers. Per NFS (1976), these two sources resulted 

in collective dose for maintenance workers of about 30 person-rem/yr from the VWR. However, this 

value likely reflects exposures after the HEV system came on line in October 1970. Personnel 

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exposures before the HEV being in operation would have been higher. The air washer was 

decontaminated after the plant shut down in 1972. 

In 2001 the dose rate on the ventilation duct in the VWR upstream of the inlet plenum was 1.5 rem/hr. 

The average for the entire length of the duct from the VWR to the VEC averaged 34 mrem/hr at 6 in.

with a maximum of 130 mrem/hr (WVDP 2005). 

A51.1.1

The working area exposure rate for a pump changeout in the VWR was in October 1969 was 10 R/hr 

to 15 R/hr. 

External Exposure Data

1983 estimates of exposure rates in the VWR were general levels of 200 mR/hr to 1,000 mR/hr and in 

excess of 5 R/hr at the washer itself (WVNSC 1994, p. 85). In 2001 the dose rate on the floor ranged 

from 10 mrem/hr to 1,200 mrem/hr gamma. The highest dose rate was in an area around the floor 

drain that read 2,500 mrem/hr beta-gamma (WVDP 2005). 

A51.1.2

An air sample taken in the VWR on October 23, 1967, showed 4E-10 μCi/cm3 of “long-lived alpha” 

and 3E-10 μCi/cm3 of “long-lived beta” (Wenstrand 1967d, p. 1). 

Airborne Contamination Data

A52.0 WASTE TANK FARM

The Waste Tank Farm (WTF) area included the underground tanks used to contain high-level liquid 

wastes from fuel reprocessing. There were separate tanks for PUREX waste (neutralized) and 

THOREX wastes (not neutralized). The PUREX tank was 750,000 gal. The THOREX tank was 

15,000 gal. Separation and solidification of these high-level liquid wastes were a major focus of the 

WVDP. 

A52.1 RADIOLOGICAL CONDITIONS

A52.1.1

An April 19, 1973, memorandum (Wenstrand 1973, p. 2) describes “major problems” associated with 

radiation levels at the WTF shelter. It says the exposure rates make the area a high radiation area, so 

it must be locked. The sources of the exposure rates were vent lines, piping, and resin units.

External Exposure Data

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ATTACHMENT 

B

LANDAUER EXPOSURE REPORT CODES

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